ABSTRACT
Year : 2023 | Volume
: 46 | Issue : 5 | Page : 163--229
Theme 4. Radiation Dosimetry (External, Internal and Biological)
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Abstract - 41101: Indoor radiation monitoring using thermoluminescent dosimeters in Shahbag Thana, Dhaka city
S. Munira, M. S. Rahman1, S. Yeasmin1, M. K. U. Sikder
Department of Physics, Jahangirnagar University, Savar, 1Health Physics Division, Atomic Energy Centre Dhaka, Dhaka, Bangladesh
E-mail: [email protected]
Introduction: Indoor radiation monitoring is more important than outdoor as people spend most of the time at indoor places.[1] Shahbag Thana is one of the busiest area of Dhaka city with two hospitals (BSMMU and DMCH), two public universities (DU & BUET) and Atomic Energy Centre Dhaka (AECD). Various kinds of radioactive materials and radiation generating equipment are being used in these hospitals for diagnostic and therapeutic purposes. AECD is a radiological facility where different types of radioactive materials are used for non-destructive testing and radiation generating equipment are used for service, training, and for Research & Development purposes. In view of the above indoor radiation monitoring study has been carried out in Shahbag Thana area.
Materials and Methods: Indoor radiation monitoring was carried out in the Shahbag Thana during September-December 2021. Twenty indoor locations were selected for monitoring radiation with a monitoring frequency was 30 days. TLDs were deployed at safe position in the rooms at 1 meter above the ground. The distance of TLDs from the side wall of the rooms were varied from 1 meter to 5 meters. All houses were made using building materials such as bricks, concretes, sands, etc. Thermoluminescent Dosimeters (TLD) were read out using Harshaw TLD Reader and the reader is operating by WinREMS software.
Results and Discussion: Mean indoor radiation dose for 30 days in the study area was found to be 253.3 ± 22.8 μSv. The measured dose for 30 days ranged from 208.1 ± 86.2 μSv to 289.8 ± 124.0 μSv. The measured dose range were observed to be higher in comparison with other areas.[2] The measured dose in the present study could be due to natural radioactivity of building materials. The data generated in the study would serve as baseline data in the area. Also, a detailed study needs to be carried out to understand the impact of radiation facilities, if any, in Shahbag Thana area. This work was funded by the Ministry of Science and Technology under grant No. 572 MS.{Figure 1}
Keywords: Indoor, public, radiation, risk, TLD
References
Hashemi M, et al. Radiat Prot Dosim 2019;184:148-54.Zarghani H, et al. J Health Sci 2017;9:1-4.
Abstract - 41127: LiAlO2: Gd as potential personnel neutron dosimeter
Pragya R. Jopat1,2, D. S. Sisodiya1,2, Shashwati Sen1,2, M. S. Kulkarni1,3
1Homi Bhabha National Institute, 2Technical Physics Division, BARC, 3Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
At present, personnel monitoring is mainly performed by CaSO4:Dy based TLD's in India for beta/gamma. For neutron, CR-39 SSNTD (Solid State Nuclear Track Detection) is being used. However, the response of CR-39 is very poor for the slow neutron energy regions and the background tracks increases over the period of time. In addition, the dosimetric grade CR-39 is still being imported in India. Keeping these shortcomings in mind, the present work focuses on studying lithium aluminate (LiAlO2) for potential application as luminescence based neutron dosimeter.[1] The utilisation of 6Li in the host matrix is advantageous due to high thermal neutron absorption cross section. Further to enhance the neutron sensitivity, LiAlO2 is doped with Gd and investigated for neutron detection. The material was prepared by solid state sintering technique by heating up to 1000 °C. The synthesized material was irradiated with a 90Sr/90Y beta and Am-Be neutron source (dose 20 mSv) and dose response was studied in thermally stimulated luminescence (TL) and optically stimulated luminescence (OSL) mode. The neutron response is calculated by irradiating the sample with Am-Be source keeping it inside and outside a 1 mm thick Cd sheet chosen so that there is no contribution from neutron. The response (TL/OSL) measured with Cd foil was subtracted from the response measured without Cd foil. When the sample is kept inside Cd foil the measured response is from gamma only. All the data plotted is normalized with respect to mass to have a uniform comparison. LiAlO2:Gd was found to give both TL and OSL response when irradiated with beta and neutron sources. The intensity of the response was found to increase with Gd doping as compared to pure LiAlO2. This may be attributed to the presence of both Li and Gd in the matrix which acts as neutron sensitive material.[2] Experiments for beta irradiation were carried out with samples having different amount of Gd doping. The best response was measured for sample with 0.5% of Gd doping. All neutron measurement was carried out for 0.5% doped LiAlO2.
In the TL measurements two glow peaks are observed at 150°C and 330°C for both beta irradiation and neutron irradiation but the intensity of peak at 150°C is very less for neutron irradiation [Figure 1] and [Figure 2]. As low temperature peak has the tendency to fade at room temperature the 330°C peak can be used for dosimetric purpose. The presence of single glow peak is always beneficial for a TL dosimeter. In case of OSL measurements on neutron irradiation, it is clearly observed that Gd doped LiAlO2 is more sensitive than pure LiAlO2 [Figure 3]. Thus our studies concludes that Gd doped LiAlO2 can be investigated as a potential personnel neutron dosimeter. Further studies on dose linearity, fading, MDD and comparison with standard dosimeters are underway. With proper studies this material can be used for personal dosimetry in mixed field applications.{Figure 2}{Figure 3}{Figure 4}
Keywords: Aluminates, dosimeter, luminescence
References
Kellerman DG, Kalinkin MO, Akulov DA, Abashev RM, Zubkov VG, Surdo AI, et al. J Mater Chem C 2021;9:11272-83.Ismail SS, Sani SA, Khandaker MU, Tamchek N, Ridzuan CM, Karim JA, et al. Radiat Phys Chem 2021;188:109654.
Abstract - 41151: Energy verifications of linear electron accelerator: important criteria for safety and beneficial applications
Nishant Chaudhary1,2, D. Bhattacharjee1,2, R. B. Chavan1, R. R. Tiwari1, U. M. Yerge1, P. C. Saroj1, A. Sharma1
1Electron Beam Centre, Accelerator and Pulse Power Division, BARC, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Electron accelerators have established its beneficial importance in multidisciplinary fields including food, agriculture, medical, radiation therapy, polymers, semiconductors, waste treatments, security, radiaography and basic research. Other than basic research and radiation therapy related uses, the regulatory limit of energy for electron accelerator is limited up to 10 MeV to avoid / minimize any chance of neutron generation and delayed radioactivity. Additionally the radiation safety outlook and shielding design of accelerator are governed by electron beam (EB) energy. In case of RF Linear accelerators (LINAC), EB energy is a dynamic parameter which is decided by combination of forward RF peak power and peak beam current. Therefore accurate measurement of EB energy is necessary for the intended application and to ensure no radioactivity generation in the radiation processed product. Presently we are reporting EB energy verification of 10 MeV LINAC, at Electron Beam Centre (EBC), Kharghar, Navi Mumbai, through two independent methods: (i) Current Ratio method and (ii) Dosimetry method.
(i) Current Ratio Method: Two plates of aluminum were taken and arranged as top and bottom plates with separation of 10 mm. The thickness of top and bottom plates were 12 mm and 25 mm respectively. The set up was kept below the beam exit window. The current constituted in each of the plates was tapped, using oscilloscope, in form of voltage built across a known resistance. The ratio of current constituted in top plate (IT) to the added current constituted in both plates (IT + IB) concludes most probable energy of EB.[1] These ratios have been generated for different combination of forward RF peak power and peak beam current and summarized in [Table 1]. (ii) Dosimetry Method: The range of electron beam in a particular material is proportional to energy and inversely related to density of material.[2],[3] These are correlated with following equation:{Figure 5}{Table 1}
Range = (0.524 E - 0.133)/density (1)
where range is in 'cm'; density is in 'g/cm3' and E is most probable energy in 'MeV'. Particularly for aluminum, the most probable energy of EB is related with range as follows.[4]
Ep (MeV) = 0.20 + 5.09 RP (cm) (2)
Thus the energy of EB is practically measured by determining the depth–dose distribution along the beam axis. Aluminum plates (each of thickness around 1 mm) were taken and radio-chromic film dosimeters were inserted between two plates such that no air gap remained. Resulted depth-dose profile shown in [Figure 1]. Full penetration depth is 19.2 mm which concludes most probable EB energy as 10.1 MeV and 9.97 MeV as per relation (1) and (2) respectively.[2],[4] The observed variation is around 0.3 % which is almost negligible. Two independent techniques for EB energy verification of LINAC have been explored. Both the techniques show quite a good aggrement of 10 MeV EB energy. The exercise helps the operator to set the RF peak power and peak beam current combination to get fixed EB energy for desirous applications and to meet the regulatory mandate also.
Keywords: Depth dose, dosimetry, LINAC, radio-chromic films, two-plate method
References
Fuochi PG, Lavalle M., Martelli A, et al. Radiat Phys Chem 2003;67:593-8.Kalia V, Joseph D. Encyclopedia of Nuclear Energy. Vol. 4. Elsevierl; 2021. p. 224-35.Chaudhary N, Sharma A. External Report BARC 2020/E/004; 2020.ISO/ASTM 51649; 2015(E). p. 24.
Abstract - 41160: Dose evaluation of the caregiver of X ray CT examination
Hiroki Ohtani, Akari Hirata, Aimi Ishikawa1
Department of Radiological Technology, Teikyou University, Tokyo, Japan
E-mail: [email protected]
Introduction: The frequency of X ray CT examination is high, and exposure of a patient tends to become high. Moreover, X ray CT examination is carried out also in emergency care, and a caregiver need to be present at the time of CT scan. In this case, although a radiologic technologist may care for at an X ray CT room, the exposure management as a caregiver other than occupation exposure is needed. However, there are few reports of dose evaluation of the caregiver at the time of X ray CT examination. The purpose of this research is to measure dose to the caregiver using a human phantom at an appropriate distance. X ray CT equipment is ECLOS by a Hitachi medical company [Figure 1]. The human body phantom as a caregiver placed at a distance of 1 m from the centre of CT gantry and close enough to couch of the CT machine. The fluor glass dosimeter was stuck on the surface of a caregiver phantom. A fluor glass dosimeter is Dose Ace GD-301 by an Asahi techno glass company [Figure 2]. Measurement parts are a eye lens, fingers, a chest, and a genital gland. Head X-rays CT photography and body part X-rays CT photography were performed using the human body phantom as a patient [Figure 3]. Head X-rays CT photography is a single scan, and CTDIvol is 59.6 mGy. Body part X-rays CT photography is a helical scan, and CTDIvol is 15.0 mGy. Each condition is described in [Table 1].{Figure 6}{Figure 7}{Figure 8}{Table 2}
Results and Discussion: A caregiver's position is 1 m from the center of CT gantry. The dose of the caregiver when X-rays CT of a head and the body is shown in [Table 2]. Fingers showed the biggest value at body CT is 42960 microGy. This is because those fingers of caregiver are in a gantry. The dose of chest and gonad are larger than the eye lens.{Table 3}
Because of position to gantry, Chest and gonad is near center of gantry and eye lens is far of gantry. About Head CT, it was a single scan, and since Body CT was a helical scan, in the dose, a helical scan became high.
When caring for, in order to support a patient by one's hand, it is not avoided that the dose of fingers becomes large. However , it is possible to reduce the dose to the caregiver with proper positioning with respect to CT gantry. Moreover it seems that a caregiver's dose decreases by wearing protective clothes for radiation such as lead glove etc. to reduce the finger dose.
Conclusion: The caregiver's dose in X ray CT examination was calculated in the human body phantom experiment. The dose changed with parts of a caregiver's body and the position at the time of care and the influence of a way to support were great. Moreover, it was suggested by devising the method of protecting to each part that a dose decreases.
Keywords: Caregiver dose, glass dosimeter, X ray CT
Abstract - 41195: Comparative study for estimation of uranium using UTEVA and DGA resin in urine samples
Soumitra Panda, Smita Thakur, J. R. Yadav, P. D. Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Uranium is a naturally occurring element found in the earth's crust. Drinking water and food are the main source of uranium intake for general population. Intake of natural, depleted, or low enriched uranium compounds possesses the risk of systemic deposition. For soluble U compounds, chemical toxicity overrides the radiological toxicity. Low level of intake estimation of uranium is generally done through urine bioassay technique. Usually, anion exchange chromatographic techniques coupled with alpha spectrometry is used for estimation of U in urine samples. Of late much emphasis is put on rapid estimation of radionuclide for better medical management of the exposed individual. In the previous study, method was standardized to estimate uranium in urine using a combination of DGA resin and LED fluorimetry technique.[1] In the present study, procedure using UTEVA resin is standardized as UTEVA resin is specific for extraction of U. The procedure developed will be useful specifically for bioassay monitoring of radiation workers handling U. The standardized method was also compared with the method previously standardized using DGA resin.
Materials and Methods: Sixteen nos. of synthetic urine samples[2] comprising 1000 mL each were prepared for the study. All the samples were spiked with known amount of U(nat). 232U tracer was also added to the samples to validate the radiochemical recovery using alpha spectrometry technique. Samples were wet oxidized using conc. HNO3 and H2O2 followed by co-precipitation of U along with Ca3(PO4)2 in alkaline medium. Ca3(PO4)2 precipitate was centrifuged and wet oxidized using conc. HNO3. After wet oxidation, samples were evaporated to dryness to obtain a white residue. This residue was then dissolved in 15 mL of 1 M Al(NO3)3 prepared in 5M HNO3 for U separation through DGA (N,N,N',N'-tetra-2 ethyl hexyl di glycol amide) resin (make Eichrom). Uranium was eluted from DGA column using 0.1M HNO3.[1] In case of chemical separation using UTEVA (Dipentyl, Pentyl phosphonate) resin (make Eichrom), Ca3(PO4)2 precipitate was dissolved in 15 mL of 1 M Al(NO3)3 prepared in 3 M HNO3 and loaded on the column and washed with 10 mL of 3 M HNO3. The column was rinsed with 5 mL of 9M HCl followed by 10 mL 0.05 M oxalic acid prepared in 5M HCl. Uranium was eluted with 25 mL of 0.01M HCl. Uranium eluted from both the resins were collected in different glass beakers and evaporated to dryness. 5mL of 0.1M HNO3 was added to the beaker and kept for few minutes. These solutions were divided into two parts., viz. 1 mL and 4 mL. 1 mL solution was used for uranium estimation using LED fluorimeter (model Quantalase LF2) in standard addition mode[3] and the remaining fraction was electrodeposited for alpha spectrometry.
Results and Discussion: The experimental results determined by alpha spectrometry for spiked synthetic urine sample is given in [Figure 1]. The average radiochemical recovery observed for DGA and UTEVA was ~85 ± 6.5%. Optimum recovery was obtained when samples were loaded in 5M HNO3 on DGA resin & 3M HNO3 for UTEVA resin. Activity measurement by both alpha spectrometry and LED fluorimetry are within ±5%. MDA of both the procedures assessed for radiochemical separation through DGA as well as UTEVA followed by LED fluorimetry is 5ng/d. Counting time of only 5 min is required with LED fluorimeter whereas to achieve same sensitivity by alpha spectrometry, sample needs to be counted for 3-4 days.{Figure 9}
Conclusions: It is observed that extraction of uranium was relatively faster using UTEVA resin as compared to DGA resin. Both, DGA as well as UTEVA resin can be used for efficient separation of uranium from urine matrix. However, added advantage of DGA resin is that it can be used for sequential separation of multiple actinides in the same sample. LED fluorimeter is rapid and more sensitive than alpha spectrometry.
Keywords: DGA, LED fluorimeter, uranium, UTEVA
References
Panda S, et al. Book of Abstract. NUCAR, 184; 2021.Wankhede SM, et al. Sep Sci Technol 2013;48:2431-5.Kumar SA, et al. J Radioanal Nucl Chem 2012;294:439-42.
Abstract - 41207: Track response in CR-39 for p+nat(Li+C) system at different proton energies
G. S. Sahoo1, S. Paul1, S. P. Tripathy1,2, A. A. Shanbhag1, S. C. Sharma3, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Nuclear Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
There is recent interest in utilizing 7Li target, which can produce a relatively high yield of quasi-mono-energetic neutrons in the forward direction via the 7Li (p, n) reaction, which has wide applications in fusion program, accelerator transmutation of waste, radiation damage studies, medical isotope production, cross section experiments, calibration of neutron measuring instruments, etc.[1] In many applications, the Li target requires a thick backing substrate viz. C, Ta, Pb etc. which may affect the neutron yield. Hence in this work, neutron yield has been determined for the p+nat(Li+C) system where natLi target (300 μm) was attached to thick natC backing (7 mm) and bombarded with protons of 4 different energies, viz. 8, 12, 16 and 20 MeV. The neutron yield was estimated using FLUKA Monte Carlo code[2] and measurements were carried out using CR-39 detectors. The irradiation of CR-39 (1.5 mm thick) was performed at 6m irradiation facility of BARC-TIFR Pelletron Linac Facility. After irradiation, the detectors were subjected to chemical etching with 6.25N NaOH at 700C for 6 hours. The neutron spectrum of p+nat(Li+C) system for 20 MeV protons obtained from the FLUKA simulation is shown in [Figure 1]. The possible contributions to the neutron generation for the present system are from various nuclear reactions on the Li and C isotopes with major contributions from 7Li and 13C. Among these, the emissions of neutrons from proton interaction with the thick natC were expected to follow the evaporation emission pattern. However, the presence of 7Be (due to 7Li (p, n) reactions), the discrete energy states were expected produce high energy neutrons. The emission of neutrons generated from the de-excitation to the ground (n0) or 1st excited state (n1) were found with a maximum energy up to 17.99 MeV and the de-excitations to the second discrete state (n2) was found with an energy of 13.31 MeV. A continuum at low energies is from the three body break up reaction of 7Li as well as 13C evaporation. Likewise, the neutron spectra of other proton energies were also obtained from the FLUKA simulations. The variation of neutron fluence with the proton energy obtained from the FLUKA simulation is shown in [Figure 2]. On analysing the neutron spectrum in case of 8 MeV incident protons, the maximum (61%) contribution was found to be from the n0 and n1 emissions. At 12 MeV, the contribution from the n0, n1 state had a sharp reduction (~17%) resulting overall decrease in the yield [Figure 2]. At higher energies, significant increase in the lower energy evaporation neutrons has led to an increase in the integral neutron emission yields. From the experiment, similar observations were found in the behavior of yield (track density) against proton energy which is presented in [Figure 2]. The track response (tracks/neutron) of CR-39 was determined from the ratio of track density and neutron fluence estimated from FLUKA. The responses were found to be 1.132×10-4, 1.290×10-4, 9.650×10-5, 7.194×10-5 tracks/neutron for 8, 12, 16 and 20 MeV protons respectively. The response obtained in the present study can be useful in estimating the neutron field around particle accelerator environment as well as for the radiation protection aspects in the facilities where p+nat(Li+C) system are being used for neutron production.{Figure 10}{Figure 11}
Keywords: CR-39, FLUKA, neutron yield, track response
References
Mashnik S, et al. Los Alamos National Laboratory Report LA-UR-00-1067, Los Alamos, NM; 2000.Bohlen T, et al. Nucl Data Sheets 2014;120:211.
Abstract - 41253: Estimation of inhalation dose from 222Rn, 220Rn and their progeny concentrations at uranium oxide facility, Kalpakkam
S. Murugan, N. Praveena, Rajababu Mari, N. Ravikumar, N. Chitra1, G. Ganesh2, M. S. Kulkarni2
Health Physics Unit, WMF, BARC(F), 1Health and Safety Laboratory, IGCAR, Kalpakkam, Tamil Nadu, 2Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Inhalation of 222Rn(Radon) and 220Rn(Thoron) are responsible for more than 50% of the radiation dose humanity. Natural decay of 238U produces 226Ra, which intern undergoes radioactive decay gives the radioactive gas 222Rn. 238U is generally found in radioactive equilibrium with its daughter product 226Ra. Being radioactive gases, 222Rn and 220Rn will try to emanate from the solid material to the environment immediately after production. In the present study 222Rn and 220Rn and their progenies concentrations were measured at different location of uranium oxide facility (UOF). Using the dosimetric approach, the inhalation dose due to this radionuclides have been estimated. Air samples to monitor the natural radioactivity of uranium daughter products has been collected from the different areas of the UOF. The concentration of 222Rn and 220Rn were measured using pin-hole dosimeter which has two compartments separated by a central pin-holes disc made up of HDPE material, acting as 220Rn discriminator. LR-115 films are fixed in both the compartments. 222Rn and 220Rn progeny concentration were measured using direct radon progeny sensors (DRPS) and direct thoron progeny sensors (DTPS). Direct progeny sensing detector system is based on selectively registering alpha tracks originating from the deposited progeny activity on LR-115 type solid-state nuclear track detectors. The selection of alpha particle energies is achieved by mounting absorbers of suitable thicknesses on the LR-115 detectors. The measurement was carried out during the period of April to June 2021.The dosimeters and the film badges used for this study were indigenously developed by BARC, and calibrated against the standard source of 222Rn gas (Model RN -1025, Source activity of 110.6 kBq as on March 1996). The measured 222Rn, 220Rn and their progeny concentrations are presented in the [Table 1]. The concentration of 222Rn and 220Rn varies from 13 ± 3 to 78 ± 4 Bq/m3and 20 ± 4 to 68 ± 8 Bq/m3 with the average value of 38.72 ± 5.4 Bq/m3 and 39.36 ± 6.4 Bq/m3 respectively. The EERC and EETC values varied from 1.30 ± 0.4 to 23.72 ± 1.6 Bq/m3 and 0.51 ± 0.1 to 3.11 ± 0.2 Bq/m3 respectively. The variation in the concentration depends upon the factors like concentration of the parent nuclides and the ventilation pattern. From the 222Rn, 220Rn concentration and EERC, EETC values inhalation dose to the radiation workers due to this radionuclides were estimated. The annual inhalation dose varies from 0.06 to 0.67 mSv/yr with the average value of 0.32 mSv/yr. The average concentration of 222Rn is lesser than the global average of 42 Bq/m3 and the average 220Rn concentration is higher than the global average in the UOF. This is due to 50 - 60 ppm of 232Th impurity in the PHWR fuel, converts into 220Rn after irradiation. The average inhalation exposure is 0.32 mSv/yr, which is 1.6% of ICRP annual dose limit.{Table 4}
Keywords: Inhalation dose, radon, thoron
References
Sahoo BK, Sapra BK, Kanse SD, Gaware JJ, Mayya YS. Radiat Meas 2013;58:52-60.Umash Reddy K, Kaliprasad CS, Suresh C, Ningappa C, Beena Ullala Mata BN, Srinivasa E. Radiat Prot Environ 2021;44:146-51.
Abstract - 41274: Gamma spectrometry of KAMINI reactor coolant lines
T. K. Srinivasan, P. M. Annadurai, R. Akila, R. Sarangapani
Health Physics Section, Health and Industrial Safety Division, Safety, Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
KAMINI is a U233 fuelled tank type, beryllium oxide reflected, de-mineralized light water cooled research reactor. The nominal reactor power is 30 kWt and uses a low fuel inventory of ~0.6kg of 233U and 83g of 239Pu in the form of U-Al and Pu-Al alloy plates. Cooling of the reactor is by natural convection and an external heat exchanger cools the tank water to control the inlet temperature during reactor operation. The impurities / corrosion activation products produced migrate and undergo different mechanisms[1] to concentrate at the pipe line bends and contribute to increase in ambient radiation level and dose consumption. The gamma spectrometry of the coolant lines of the reactor provides information about the deposited impurities and corrosion products. Prior to gamma spectrometry, the gamma radiation level on the coolant lines were measured using hand held digital teletector and scintillometer. The selection of locations on the coolant line for gamma spectrometry was made based on the radiation survey carried out during reactor operation and shutdown. Efficiency calibration was carried out for NaI) based RIIDEYE-X and portable HPGe by simulating the coolant line geometry and detector arrangement. [Figure 1] shows the Spectrum obtained using RIIDEYE near Reactor vault door 20 kW(t). [Table 1] shows the nuclides present in the coolant during reactor operation. The gamma radiation level observed during reactor power on these spots is due to the FPNGs and their short lived radionuclides 88Rb and 138Cs generated. 133Xe and 135Xe could be identified only by RIIDEYE-X on the coolant lines inside the HX room top during power operation as the sensitivity at 81 kev and 250 kev is higher than CZT. Significant decrease in the radiation filed as well as in the count rates between 100kev to 1 MeV region during reactor shutdown. Coolant line spectra results did not indicate the presence of any corrosion activation products such as 60Co and fission products 137Cs - implying no observable internal contamination. 137Cs as well as traceable amount of 60Co were identified in the mixed bed and in the regenerated water. The presence of 137Cs could be attributed due to the tramp fuel fissions.{Figure 12}{Table 5}
Keywords: Coolant, efficiency, gamma, spectrometry
Reference
Trojanowicz M, Kołacińska K, Grate JW. A review of flow analysis methods for determination of radionuclides in nuclear wastes and nuclear reactor coolants. Talanta 2018;183:70-82.
Abstract - 41285: Patient organ dose estimation due to some selected fluoroscopy procedures using kerma area product meter: The Ghananian experience
Edward Gyasi, Mary Boadu, Cyril Schandorf, Prince Kwabena Gyekye
E-mail: [email protected]
Aim: With the acquisition of the Kerma-Area-Product (KAP) meter, patient organ doses were estimated in order to analyze patient dose trends due to fluoroscopy exposure in two fluoroscopy centers. This gave the opportunity to report patient doses due to fluoroscopy exposure using the appropriate dosimetry procedure.
Study Design: Cross-sectional study.
Place and Duration of Study: Two fluoroscopy machines located in Greater Accra Region of Ghana in Korle-Bu Teaching Hospital and Cocoa Clinic. The duration of the study was within six and a half months.
Methodology: 182 adult patients undergoing barium enema, barium meal, barium swallow, myelogram, hysterosalpingography and urethrogram examinations collectively were investigated (98 men, 84 women, age group 20-81 years). Radiation dose was measured using KAP meter. The KAP readings, patient's data and other relevant information from the control console were used to estimate organ doses using Monte Carlo base program (PCXMC version 2.0). Quality control tests were performed on the two fluoroscopy machines before the start of the study to ensure that they were performing self-consistent with national and international requirement.
Results: The ovaries, breast, thyroid and testes recorded high doses for barium enema, barium meal, barium swallow and retrograde urethrogram examination respectively. Mean KAP values measured were 23.57 ± 1.78 Gy.cm2, 18.08 ± 2.08 Gy.cm2, 5.99 ± 0.62 Gy.cm2, 8.53 ± 0.67 Gy.cm2, 2.13 ± 0.15, Gy.cm2 and 1.47 ± 0.07 Gy.cm2 for barium enema, barium meal, barium swallow, myelogram, hysterosalpingography and urethrogram examinations respectively.
Conclusion: The recorded KAP values for all the examinations were compatible with ICRP values but in some cases where a little bit lower. The KAP values were also higher than NRPBs' values except for barium swallow examination which was comparable. Due to the varying patient doses in the institutions, standard protocol for fluoroscopy procedure is recommended.
Keywords: Fluoroscopy, kerma-area-product, Monte Carlo, organ dose, radiography
Abstract - 41331: Investigation of utilization of indirect beam in cave room of 7 MeV LINAC
J. K. Divkar, M. A. Toley1, S. J. Shinde1
Radiation Safety Systems Division, Bhabha Atomic Research Centre, 1Radiation and Photo Chemistry Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: LINAC is a 7 MeV electron accelerator[1] used as a pulsed radiolysis facility which can be operated in single (25 ns to 2 μSec) and multiple (2 μSec) pulses with variable Electron Beam (EB) dose. It was intended to assess primarily the levels of radiation doses arising during operation of the facility inside cave room. The radiation doses at various positions was measured using ionization chamber-based dosimeters (DRDs) at various angles and distances.
Materials and Methods: In first phase, the measurements were performed at ten (One DRD each) different positions (random on platform) except trajectory height in cave room for repetitive pulses. In second phase, the measurements were performed at various angles of the trajectory (0° and 45° at trajectory height) for single pulse. The measurements were performed “Sample In” condition for 2 μSec pulse.
Results and Discussions: The radiation doses recorded in DRDs along with EB dose in target sample (measured by kinetic absorption spectrophotometric technique) for different pulse duration are given in [Table 1]. The doses are ranging from 0 to 36 mGy. The higher doses were observed.{Table 6}
Conclusions: The single pulse resulted in high doses at trajectory height. The wide range of doses offers more space for studies in various fields. The chemical/luminescence dosimetry methods can be exercised for generation of detailed investigation at high EB dose.
Keywords: Angles, beam, electron accelerator, radiation doses
Reference
Manual of the Radiation Dynamics Super X Linear Accelerator for Pulse Radiolysis, Radiation Dynamics Ltd., U.K.
Abstract - 41340: Correlation of workplace monitoring with exposure profile of workers
S. K. Nayak1, V. Ramprasath1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: A nuclear facility handling plutonium is designed based on the defense in depth principle, which includes several barriers between the source and the environment.[1] In addition to this engineered safety features, administrative controls are implemented for safe operations. In a mixed field of gamma and neutron radiations, individual radiation workers working in the area receive both gamma and neutron doses [Hp(10)].[2] In this study, an effort has been made to correlate the exposure profile of the radiation workers with their workplace monitoring in a plutonium handling facility.
Materials and Methods: Workplace monitoring: Gamma dose rates were measured in the working areas of the lab and on the product containers by gamma survey meter such as teletector (model no: Automess 6112B, manufactured by Southern Scientific Ltd, UK, dose rate range: 1 μSv/h to 10 Sv/h for gamma energy up to 3 MeV). Similarly, neutron dose rates were measured by neutron survey meters such as roentgen equivalent man (REM) counter (model no: Ludlum 2241–4, dose rate range: 1 μSv/h to 100 mSv/h for all neutron energy that is, from 0.025 to 15 Mev. Personnel monitoring: Personnel monitoring of the workers involved in the reconversion process was carried out for both gamma and neutron doses. A Thermo Luminescence Dosimeter (TLD) consisting of three Teflon embedded discs of CaSO4:Dy sealed in polyethylene pouch put in a plastic cassette was worn on the chest of the worker for measuring the gamma dose. Similarly, CR-39 foil procured from M/s Pershore Moldings, UK of dimension 3 cm × 3 cm × 0.625 mm (thickness) packed in triple laminated aluminized pouch were worn on the chest of workers for measuring the fast neutron dose.
Results and Discussion:
Dose Ratio: The persons working in the Plutonium handling glove box area received both gamma as well as neutron doses. The cumulative and average individual gamma to neutron dose ratio was found to be 9.72 and 9.43 respectively. To avoid the error in the calculation of dose ratio, the data for a period of five years have been into consideration for calculations. The errors in measurements are within ±25%.
Dose Rate Ratio: The ratio of gamma to neutron dose rates in the working area of the Plutonium handling Glove boxes and Pu oxide containers were found to be 9 & 10 respectively. Again, the dose rate data for a period of five years have been taken for calculation to minimize error. The dose and dose rate ratios observed are presented in [Table 1] and the plot of the data are presented in the [Figure 1].{Table 7}{Figure 13}
Conclusion: It is observed from the analysis of data that the ratio of gamma to neutron dose/dose rate is about 10. The factor holds good for cumulative dose of workers, for average individual dose, for glove box dose rates and for the dose rates of PuO2 containers. It can therefore be concluded that exposure profiles of radiation workers and the work place are of similar trend in a plutonium handling facility.
Keywords: Gamma and neutron radiations, spontaneous fission, personnel and workplace monitoring.
References
IAEA. Safe Handling and Storage of Plutonium. IAEA Safety Reports Series No. 9; 1988.Endo K, Momose T, Furuta S. Radiation protection at nuclear fuel cycle facilities. Radiat Prot Dosimetry 2011;146:119-22.
Abstract - 41343: Performance comparison of neutron survey meters using different neutron sources
S. K. Nayak, B. C. Muduli, Rakesh Tiwari1, Vikram Singh, B. B. Murmu, Ashok Kumar, Prem Chand1, U. P. Shriwastawa1, V. Ramprasath, G. Ganesh
Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, 1INRP(K), NRB, BARC Faciliteis, Kalpakkam, Tamil nadu, India
E-mail: [email protected]
Introduction: Measurement of neutron equivalent dose rate; Hn*(10) in the mixed field of gamma and neutron needs a high sensitive neutron survey meter with better gamma rejection factor. Traditionally BF3 or 3He proportional counters are used to measure neutron equivalent dose rates. In this study, a comparison of the performances of two different survey meters is done using two different neutron sources.
Materials and Methods:
Measuring Devices: Neutron dose rates were measured by neutron survey meters such as roentgen equivalent man (REM) counter (model no: Ludlum 2241–4, dose rate range: 1 μSv/h to 100 mSv/h for all neutron energy that is, from 0.025eV to 15 MeV and portable neutron survey meter (model no: NSN3, dose rate range: 0.1 μSv/h to 99.99 mSv/h for neutron energy from 0.025 eV to 15 MeV. The working principle of Ludlum 2241-4 is thermalization of fast neutrons and detection of thermal neutron by 3He(n, p)3H reaction whereas the working principle of NSN3 is based on two reactions; the fast neutrons are detected by H(n, n!)H and thermal neutrons by 14N(n, p)14C reaction.[1]
Neutron Sources: The sources used for checking the performances of the detectors are Pu-Be (source strength: 9.4E+5 n/s) and PuO2 (source strength: 7.63E+5 n/s). The energy of neutrons emitted by both the sources are of fast neutron energy up to 15 MeV with average energy of 2-3 MeV. The neutron emission from these sources are due to spontaneous fission and (α, n) reaction such as 9Be (α, n) 12C & 18O (α, n) 21Ne.These sources are kept at a height of 1m from the ground and 1m away from the wall of the room to avoid the reflection of neutrons during the dose rate measurement.{Figure 14}
Theoretical Calculations: The neutron sources PuO2 and Pu-Be were simulated using Monte Carlo simulation and dose rates were estimated by using ICRP-74 neutron flux to dose conversion factor in the simulation.
Results and Discussion: The theoretical and measured values of neutron dose rates Hn*(10) for PuO2 as well as Pu-Be sources are presented in the [Table 1]. The large variation of dose rate measured by Ludlum survey meter from the theoretical value for on contact of the source may be due to anisotropy of the source to detector geometry. These differences are minimum at larger distances from the sources. The dose rates measured by NSN3 survey meter were observed to be slightly higher than that of Ludlum. But the readings of both the survey meters (Ludlum & NSN3) are comparable to the theoretically calculated values.{Table 8}
Conclusion: The performance of both the neutron survey meters is satisfactory within an error of ±30%. Since the NSN3 survey meter is very light weight, survey of plant areas becomes less cumbersome for the health physicist.
Keywords: Fission, Monte Carlo simulation, NEUTRON
Reference
Available from: https://www.drct.com/neutron-deetection/ Fuji-NSN3-neutron-monitor.html.
Abstract - 41383: The facility of radiation standards in japan atomic energy agency recent activities with a focus on establishment of accredited testing laboratory
H. Yoshitomi, T. Tsuji, T. Fukami, S. Nishino, J. Takamine, T. Murayama, Y. Tanimura
Department of Radiation Protection, Nuclear Science Research Institute, Japan Atomic Energy Agency, Ibaraki, Japan
E-mail: [email protected]
The Facility of Radiation Standards (FRS) in the Japan Atomic Energy Agency (JAEA) is a comprehensive secondary standard dosimetry laboratories (SSDL). Its X, gamma-ray, beta-ray and neutron calibration fields has served for calibration and testing of radiation protection monitoring instruments over fourth decades. We will present the present status of the FRS and recent activities, especially introduction of the first accredited testing laboratory for radiation protection monitoring instruments in Japan.
Overview of the FRS
Currently, a wide range of photon fields between 33 keV (for X-ray narrow series) and 6-7 MeV (for R-F high energy gamma ray) are available using an X-ray apparatus, gamma-ray sources and an accelerator. Two different types of beta-ray calibration fields, the Beta Secondary Standard type 2 (BSS2) and the so-called JAEA Beta irradiation System (JBS) has served for calibration and testing. The available mean beta particle energy ranges from 0.06 MeV (corresponding to 147Pm) to 0.8 MeV (corresponding to 90Sr/90Y). With regard to the neutron calibration fields, the FRS has provided RI fast neutron calibration fields (252Cf and 241Am-Be), moderated neutron fields (D2O-moderated 252Cf and graphite moderated 241Am-Be as workplace simulated field), a thermal neutron field and ten different energies of mono-energetic neutron calibration fields with energies between 8 keV and 19 MeV. These fields are widely open to not only inside JAEA but also other users such as industry and academia. Consequently, more than 30,000 dosemeters are calibrated or type-tested in a single year.Recent activities
First accredited laboratory with regard to the JIS (Japanese Industrial Standard) energy performance testing for radiation protection monitoring instruments in Japan
The FRS has established a testing laboratory to test the performance of radiation protection monitoring instruments in terms of radiation energy based on the following four JISs.
(i) JIS Z 4345: Passive integrating dosimetry systems for personal and environmental monitoring of photon and beta radiationJIS Z 4416:Solid state nuclear track dosemeter for personal neutron monitoringJIS Z 4333:Portable ambient and/or directional dose equivalent (rate) meters and/or monitors for X, gamma and beta radiation JIS Z 4341:Neutron ambient dose equivalent (rate) meters
Using above several fields, we has developed the procedures to conduct the tests in the way that fully comply with the requirements of these JISs and established a quality management system to consistently produce valid results according to ISO/IEC 17025:2017. Consequently, the FRS was firstly accredited as JIS testing laboratory for radiation protection monitoring instruments in Japan under the JNLA (Japan National Laboratory Accreditation System) program based on the JIS Law on 23th June 2022. However, the certification is currently valid only domestically. Therefore, we are preparing to be certificated as as ILAC-MRA accredited laboratory as well as to expand the scope to the other performance test such as angular dependence.
Development of gamma-ray reference calibration filed in 350 keV region using 133Ba source
The photon response of a dosemeter for the energy region around 350 keV is important especially for the emergency environmental monitoring when the major contribution of dose rate would come from 131I (365 keV). Therefore, we developed the low-dose rate (~1 μSv/h) calibration field using 133Ba gamma source. The fluence mean energy was 333 keV.Calibration field corresponding to a new operational quantities proposed in ICRU95
The ICRU has introduced a new definition of the operational quantities in ICRU Report 95. SSDL use the operational quantities to calibrate or test the dosemeters. We evaluated the conversion coefficients from measurement standards to the new operational quantities for some calibration fields of the FRS. This would make possible to investigate the impact on the characteristics of the dosemeters when introducing the new quantities.
Keywords: Accredited testing laboratory, calibration field, new operational quantity, SSDL
Abstract - 41385: Investigation of the impact of introducing ICRU 95 operational quantities on photon dosimetry based on photon spectra in nuclear industry
H. Yoshitomi, T. Tsuji, T. Fukami, S. Nishino, Y. Tanimura
Department of Radiation Protection, Nuclear Science Research Institute, Japan Atomic Energy Agency, Ibaraki, Japan
E-mail: [email protected]
Radiation monitoring for external radiation protection is made using a set of operational quantities. The ICRU has proposed to change the definitions of the operational quantities in ICRU Report 95. This study investigated the impact of this change on the radiation dose measurements at workplaces in nuclear industry from the aspect of the photon spectra. The proposed operational quantities are defined as the products of particle fluence or air kerma and a conversion coefficient which is a function of energy of radiation of interesting. The conversion coefficients for mono-energetic radiations are tabulated in ICRU Report 95. Their numerical values and energy dependence are different from those of the existing ICRU 39/51 conversion coefficients. With regard to photons, the differences between the ICRU 95 and the ICRU 39/51 conversion coefficients for ambient dose and personal dose are significant for low energy in particular. In nuclear industry, radionuclides which emit gamma-rays with energies more than 500 keV in their decay such as 137Cs, 54Mn and 60Co are frequently present. However, scattering due to shielding or any surrounding materials can increase fraction of lower energy photons at a point where radiation doses are measured. Therefore, it is important to obtain photon energy spectra in workplaces to discuss the impact of the changes of the operational quantities. Eight workplaces were selected from nuclear research facilities, outside of the reactor buildings being decommissioned at the Fukushima Daiichi nuclear power plant and a nuclear power plant under maintenance. More than thirty pulse height distributions acquired by a CdZnTe detector, LaBr3(Ce), or NaI(Tl) scintillators at the eight workplaces were unfolded to obtain photon energy spectra[1] [Figure 1]. The spectra were then multiplied by ICRU 95 and the ICRU 39/51 conversion coefficients to derive the operational quantities. Corresponding operational quantities such as ambient dose, H*, and ambient dose equivalent, H*(10) were compared. The results indicated that the new operational quantities used for whole body monitoring are smaller than the corresponding existing ones by around 16 %. However, this ratio did not vary substantially between higher energy and lower energy of the scattered photons [Figure 2]. This indicates that dose measurements in terms of ICRU 95 operational quantity could be conducted appropriately by using existing dosemeters when they are properly calibrated with the new quantity and having good energy dependence in terms of existing quantity. With regard to the quantities for eye lens monitoring and extremity monitoring, no significant difference can be found between the new and existing operational quantities. This study was supported by the Nuclear Regulation Authority, Japan.{Figure 15}{Figure 16}
Keywords: Dosimetry, ICRU95, operational quantity, photon spectrum
Reference
Tanimura Y, et al. Prog Nucl Sci Technol 2019;6:134-8.
Abstract - 41449: Chemical induced premature chromosome condensation technique for biodosimetry
D. Bakkiam, A. Arulananthakumar, Swetha Sonwani, O. Annalakshmi, C. V. Srinivas, B. Venkatraman
RDS, EAD, SQRMG, IGCAR, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: Chromosomal aberrations are considered to be most sensitive and specific indicators of ionizing radiation (IR) induced genetic damage in human peripheral blood lymphocytes (PBLs). Scoring of dicentric chromosome aberrations in PBLs is a simple and established technique in determining suspected over exposures. In case of an exposure to IR over 5 Gy, the conventional dicentric assay (DC) does not provide accurate dose measurement due to three reasons i.e. (i) lymphopenia at high doses (ii) radiation induced cell cycle arrest resulting in low mitotic index and (iii) saturation of DC, thus reducing the precision of dose estimation. Premature chromosome condensation (PCC) overcomes the above mentioned problems of DC assay where use of specific inhibitors of type 1 and 2A protein phosphatases directly induces PCC at any phase of the cell cycle (G1/2 – Gap1,2; S – synthesis; M/A – metaphase anaphase transition). Lamadrid et al. 2007[1] have found that estimation of radiation dose of upto 20 Gy for low LET exposure was possible by applying chemical induced PCC assay. Through scoring PCC-rings in Giemsa stained PCC spreads, Hayata et al., 2001[2] estimated the doses in Tokaimura criticality accident victims.
Objective: As per IAEA guidelines, every biodosimetry laboratory should establish dose response curve (DRC) for PCC assay. The present work was carried out to construct a dose response curve (DRC) for Co-60 gamma radiation induced PCC -rings in blood cells exposed to a dose range of 1Gy to 20 Gy and validation of DRC by radiation dose blinded scoring. Irradiation of blood samples and culture methodology: Blood sample aliquots were irradiated in Cs-137 source installed at Calibration Facility, IGCAR and the doses ranged from 1Gy to 20Gy. Culture methodology, slide preparation and scoring criteria were as per IAEA 2011 EPR biodosimetry guidelines. Calyculin A was added to blood cultures during final hour of 48h incubation. For constructing of DRC, scoring of 500 PCC- spreads or 100 PCC-rings/dose points were done. DRC was constructed using MS-excel (Dose in Gy vs PCC – rings/cell). PCC- rings frequencies between the doses were statistically significant (t-test).
Results and Discussion: Unlike DC assay which allows visualization of chromosomes only in metaphase stage (M) of mitosis, Calyculin A induced PCC allows visualization of the chromosomes in all stages of the cell cycle (G1, S, G2 and M/A-phase). In this study, PCC-rings were scored in G2 and M/A PCC spreads [Figure 1] and DRC was constructed. Frequency of PCC – rings in control sample was very low (1 in 1000 PCC -spreads). PCC-ring frequency in irradiated samples varied from 0.02 to 1.25. DRC follows linear relationship with the dose (Y = 0.0635 D + 0.001, where D - dose and Y – ring/cell). Our results were comparable with dose response curves reported by S. Balakrishnan et al., 2010[3] and Puig et al., 2013[4] [Figure 2]. Validation of DRC by radiation dose blinded test revealed that the variation in dose estimates of samples were well within the acceptable limits (± 20%).{Figure 17}{Figure 18}
Conclusion: Our results show that PCC-ring assay by chemical method is appropriate for use as biodosimeter for very high dose exposures (upto 20Gy) of ionizing radiation. When compared to cell fusion PCC, chemical induced PCC involves less technical expertise, easier to implement and relatively easy scoring. This assay can be performed in any kind of accidental emergencies involving radiation exposures and to provide dose information to assist the physician for taking appropriate therapeutic decisions.
Keywords: Biodosimetry, Calyculin A, dose response curve, premature chromosome condensation – rings, peripheral blood lymphocytes
References
Lamadrid AI, et al. PCC-ring induction in human lymphocytes exposed to gamma and neutron irradiation. J Radiat Res 2007;48:1-6.Hayata I, et al. Cytogenetical dose estimation for 3 severely exposed patients in the JCO criticality accident in Tokaimura. J Radiat Res 2001;42:149-55.Balakrishnan S, et al. Biodosimetry for high dose accidental exposures by drug induced premature chromosome condensation (PCC) assay. Mutat Res 2010;699:11-6.Puig R, et al. Suitability of scoring PCC rings and fragments for dose assessment after high-dose exposures to ionizing radiation. Mutat Res 2013;757:1-7.
Abstract - 41471: Time dependent variation of Cs-137 surface contamination profile on stainless steel and linoleum surfaces
Clinton S. A. Fernandes, Sanjay Singh, R. K. Mishra1, M. K. Sureshkumar
1Health Physics Division, Bhabha Atomic Research Centre, 2Waste Management Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Radiological facilities deal with various equipment and packages whose surfaces may be contaminated. Radiological clearances are given based on transferrable surface contamination, estimated by a swipe sampling method. In some cases, the activity picked up on the swipe is corrected by a suitable pickup factor which depends upon the nature of the contaminant and the surface involved. Recently, an important phenomenon known as “contamination weeping” has been reported by Bennett et al.[1],[2] In the paper, authors have mentioned the reoccurrence of loose contamination on a transportation cask, that has been cleared for loose contamination below regulatory limits. The current work has been carried out to investigate this phenomenon on two surfaces namely stainless steel and linoleum and the results are presented.
Materials and Methods: In this experiment, 1 ml of 137Cs solution (3000 Bq/ml) was spread on stainless steel (SS) and linoleum (LL) tiles (3 each) within an area of 10 cm x 10 cm. After drying overnight, square shaped cotton swabs (1” x 1“), cleaned with ethanol, were used to wipe the surface in an anticlockwise manner from the outer edge towards the center, while applying a uniform pressure. Ten consecutive swipes were taken from each surface and counted using an end window GM counter (12% efficiency for 137Cs). After leaving the tiles undisturbed for 10 days, five consecutive swipes were taken and counted for radioactivity as described above.
Results and Discussion: The activity picked up by the first ten swipes taken on the first day after contamination, gradually decreases, as expected, since the amount of transferrable contamination on the surface keeps decreasing with successive swipes. After a gap of 10 days, however, instead of the swipe activity continuing to decrease further or being below the detection limit, it increases considerably. In some cases, the activity is even greater than that picked up on the very first swipe. This observation is seen in all sets of stainless steel and linoleum surfaces. Representative plots of one set are shown in [Figure 1] and [Figure 2]. It can be seen that the total loose activity on the 11th swipe, which is taken after 10 days is almost equal to that on the first swipe. This observation is consistently seen in case of both the surfaces and in all sets of measurements.{Figure 19}{Figure 20}
Conclusions: Investigations carried out showed the reoccurrence of significant levels of loose contamination on SS and LL surfaces after 10 days. On both surfaces, it is seen that the reoccurred value of loose contamination is as high as that obtained on the first swipe. Further research into the mechanisms of this process would reveal information that may help in designing a more effective decontamination procedure and surface preparations, thus reducing radiation exposure from delayed release of the contaminant, as well as eliminating the requirement of repeated decontaminations.
References
Bennett PC, Doughty DH, Chambers WB. Proceedings Conferences PATRAM'92; 1991.Bennett PC, Mason M, Paquin P. Proceedings Conferences 9th International Symposium on Packaging and Transport of Radioactive Materials; 1986.
Abstract - 41506: Neutron dose measurements using a wide range novel combination neutron dosimeter at CERN-EU high-energy reference field facility, CERN
Rupali Pal1, A. K. Bakshi1, B. K. Sapra1
Radiological Physics and Advisory Division, BARC, 1Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
There is an increase in ion accelerators up to GeV range around the world for spallation and cosmic neutron studies. It is challenging to measure the neutron fields produced in such facilities, having energies from thermal to beyond hundreds of MeV. High energy electron and synchrotron accelerators give rise to pulsed neutron field, which leads to pulse pile-up and saturation in active measurement devices.[1] Thus, it is necessary to develop robust passive detectors for measurement of neutrons from thermal to High-Energy range, capable of measurement in pulsed fields as well. A Combination Neutron Dosimeter (CND) has been developed for the measurement of neutron fields, from thermal (reactor environment) to high energies (up to 100 MeV, as encountered in high energy accelerator environment. The CR-39 based dosimeter has three elements namely; Boron doped CR-39 with 0.5% 10B (B-CR-39) for thermal neutrons; 1mm polyethylene (PE) converter for 100 keV-10 MeV; and 1 mm Zirconium (Zr) (purity > 99.8%) (Zr-CR-39) for high energy neutrons up to 200 MeV.[2] Zr has a threshold at 13 MeV for (n, 2n) reaction which lowers the neutrons of higher energy to about 1 MeV after interaction, to enable them to be registered in CR-39. This leads to enhancement in track density as compared to bare detector and total enhancement factor is generated against neutron energies. The CND, as shown in [Figure 1], utilizes B-CR-39 and CR-39 detectors (M/s TASL Ltd. UK) of dimensions 3 cm x 3 cm with thickness 625 μm; with the three elements sealed in triple laminated pouches. This study presents the response studies of the CND at CERF (CERN-EU high-energy reference field facility). The CND was irradiated at CT8 position of concrete roof shield at CERF facility for 118 h.[3] Typical values of ambient dose equivalent rates are 0.2 -0.3 nSv per PIC count over the exposure location on the concrete roof. Above the 80 cm concrete top shield, the neutron spectrum has a pronounced peak at around 70 MeV and resembles the high energy component of the radiation field created by cosmic rays at commercial flight altitudes. Therefore, these exposure locations provide wide spectrum radiation fields well suited to test detectors for high energy neutrons. Subsequent to exposure, the B-CR-39 detectors were etched for 2 h, 7 N KOH at 60°C optimised for development of alpha tracks only, hence, eliminatinating interference of proton tracks. CR-39 and Zr-CR-39 detetctors were etched for 7 h, 7 N KOH at 60°C. Registered tracks were counted under microscope (with 20X objective). The B-CR-39 was calibrated with standard thermal source and CR-39 with PE calibrated with 241Am-Be for dose linearity. Dose in Zr-CR-39 is calculated based on excess tracks in detector as compared to bare CR-39, discussed in the paper Pal et al., 2022.[2] Low energy neutrons (<10 MeV) do not have cross-section with Zr and hardly produce excess tracks in Zr-CR-39. The ambient dose equivalent H*(10) estimated by the CND is shown in [Table 1]. The H*(10) from CR-39 with PE is highly underestimated as compared to the Zr-CR-39. This is due to reduced interaction of high energy neutron with thin CR-39. The cumulative dose of the three components adds up to 7.33 mSv which is comparable to 8.2 mSv of the irradiated dose as given by PIC counts. Use of this CND could be useful for workplace monitoring and dose mapping in inaccessible areas in high energy accelerator environment with energies ranging from thermal to 100 MeV neutrons.{Figure 21}{Table 9}
Keywords: Boron doped CR-39, CR-39, Zr-CR-39, high energy neutrons
References
Verma, et al. RPE. 2013;36:160.Pal, et al. Radiat Meas 2022;153:106747.Pozzi F, Silari M. NIM A 2020;979:16447.
Abstract - 41517: Assessment of beta dose during the decontamination of heavy water system component
S. Sajin Prasad, M. T. Valvi, B. Madhumita1, Ranjith Sharma, N. K. Choudhary2
Health Physics Division, Bhabha Atomic Research Centre, 1Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 2Reactor Operations Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The primary coolant system of a research reactor contains radioactive fission, activation and corrosion products .The coolant water is circulated continuously through purification circuit consisting of micro filters and ion exchange resin cartridges, to remove the radioactive ionic impurities. The strainers are used in the coolant circuit to remove suspended particles otherwise, it may lead to higher radiation field on pipeline, valves, bends etc. These strainers are removed based on the reduction in coolant flow through them .The strainers are reusable components and one of the major sources of occupational exposure during its handling and decontamination. This paper describes the characterization of liquid effluent generated during decontamination of strainer and the estimation of beta and gamma dose rate using thermo luminescent dosimeters (TLD's). The strainers used in the primary coolant system accumulates the fission, activation and corrosion products radio nuclides over a period of time .The radiation field on the strainer depends on the primary coolant activity. These strainers are manually removed from the system and decontaminated. The removal and decontamination jobs of the strainer results in occupational exposure, transferable contamination in working areas and generation of radioactive liquid effluents. The general gamma radiation field measured on a strainer removed from the heavy water system using G.M survey meter was 100-150 mSvh-1. For routine Personnel monitoring, TLD's (based on CaSO4: Dy) is used and are processed in Personnel Monitoring Section (PMS/RPAD). 4 no. of such TLD's were attached serially around the strainer, to measure the dose rate, before and after decontamination. The beta dose rate estimated using TLD's are given in [Table 1]. For characterization of radio nuclides present in liquid effluent generated during decontamination, an HP Ge detector based gamma spectrometer having relative efficiency of 50% and resolution of 1.9 keV at 1332 keV gamma energy of 60Co was used. The sample was prepared in standard glass vials of suitable geometry for high resolution gamma ray spectrometry. The energy and efficiency calibration of the detector system was determined from gamma rays emitted over the full energy range of interest from multi nuclide radioactivity sources namely 133Ba, 137Cs and 60Co for a wide energy range from 80 keV to 1332 keV. The energy efficiency values are fitted in Log polynomial of order 3 for a wide energy range up to 1.332MeV.[1] The gamma spectrometric analysis result of the liquid effluent is given in [Figure 1]. The liquid effluent samples showed 95Nb (25.79%), 95Zr (24.06%), 144Ce/144Pr (23.51%), and 141Ce (15.62%) etc. as the major radio nuclides contributing to dose rate. The general beta to gamma dose rate ratio on the strainer was in the range from 2 to 3. The maximum gamma and beta dose rate, after the decontamination of the strainer was 1.5 mSvh-1 and 0.30 mSvh-1 respectively. The rubber gloves used by workers, during the handling of strainer gives an attenuation of 20% in reducing the beta dose rate. The decontamination of the strainer using ultra sonic cleaner has resulted in reducing the personnel exposure. It was observed that the average beta dose received by the personnel during the removal and decontamination jobs was within the regulatory dose limits prescribed by AERB.{Figure 22}{Table 10}
Keywords: Characterization, dose, extremity, thermo luminescent dosimeter
Reference
Gilmore GR. Practical Gamma-Ray Spectrometry. West Sussex: John Wiley & Sons; 2008.
Abstract - 41546: Comparative study on tritium-in-air concentration levels and committed effective dose
K. Srihari, V. Ramakrishna, G. Ganesh, M. S. Kulkarni
Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
The Heavy water (D2O) is used in the PHWRs as the moderator and coolant. Apart from the chain reaction, neutrons produced in the fission process also irradiate the Deuterium in D2O and produce tritium by (n,γ) reaction and subsequently form tritiated heavy water (DTO or T2O). There would be always a certain probability of leakage of the tritiated water or vapor by the failure of the valves, leaks at the joints or cracks in the pipes. The Tritium enters the body by inhalation, ingestion and absorption through the skin. The tritiated water is distributed uniformly in the body water after inhalation and transfer from lung to the blood. Skin absorption follows transfer to the lymph system and then to blood. The whole body tissue will get irradiated with the uptake of the tritium. Hence from the radiological safety point of view, it is required to regularly monitor the plant areas for air concentration levels (DAC) and the tritium uptake of all working personnel. In the present study, the data pertaining to the monitored tritium–in-air levels (DAC) of various plant and peripheral areas for six years was analysed. The samples were analysed by using Liquid Scintillation Counting system (HIDEX 300SL).[1] The committed effective dose was estimated from this and compared with the actual collective dose. The tritium has radiological half-life of 12.3 years and biological half-life of ~10 days. The effective half-life would be of ~ 10 days. The βmax energy is 18.6 keV and βavg energy is 5.7 keV.[2] It remains in the work area air in mostly vapour form which makes it to easily condense on cool surfaces for sampling. The condensed water is collected and analysed in LSS for estimating the concentration in air. The results are as given in [Figure 1] obtained by averaging the daily monitored readings over six years. The locations in A, B, C are of main plant active areas and remaining are from peripheral areas. Taking the assumptions such as two operators working in active area (locations A, B & C) for 1 hour each in a shift for daily routines and remaining duration in the peripheral areas during working hours, the annual inhalation rate of 2400 m3 etc… the committed effective dose based on this occupancy criterion is estimated as per the dose conversion factor(DCF)[3] (1.8×10-11 Sv/Bq). Internal dose(Sv) = X × Y × DCF ×BR × WH. X: DAC; Y: 3.0E5(Bq/m3)/DAC for Tritium; BR:Breathing rate=1.2 m3/hr; WH=workinghours: 6h/day *365 days/y. It is compared with the annual collective internal doses as given in [Table 1] which includes both operation and maintenance activities. It can be seen that the air concentration observed during the entire six years period in active areas remains <0.5 DAC and in peripherals it was < 0.05 DAC. The committed effective dose was within the regulatory limits in both estimated and measured cases. The measured collective dose was always on higher side attribute to the significant contribution of dose during maintenance activities.{Figure 23}{Table 11}
Keywords: DAC, DOSE, DTO, LSS
References
Hidex 300 SL – Automatic Liquid Scintillation Counter Hand Book.Knoll GF. Radiation Detection and Measurement. New York: John wiley & Sons, Inc; 2000.Martin JE. Physics for Radiation Protection a Hand Book. Weinheim: WILEY-VCH Verlag GmbH & Co.KGaA.
Abstract - 41549: Effect of optical stimulation temperature on the TA-OSL background
Anuj Soni, D. R. Mishra
Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: In case of TL readout, the main source of the background is infrared (IR) which is caused by heating of the sample holder at high temperatures. On the other hand, in case of OSL recorded at room temperature or at elevated temperatures i.e., TA-OSL, the background signal is expected to occur mainly due to scattering of stimulation which is superimposed on the actual signal emanating from the sample. In order to analyze the signal, one has to subtract the background signal from the total signal. But, if this background signal gets affected by the stimulation temperature then it can lead to un-reliable results as far as TA-OSL signal is concerned. In the present work, the effect of readout temperature on the OSL background has been reported.
Results and Discussions: The IR background emission was recorded, for the sake of comparison, by heating the un-irradiated Al2O3:C pallet, till a temperature of 400 °C at a heating rate of 2 °C per sec. The background emission shown in [Figure 1] can be seen to be very small in magnitude even at a highest heating temperature of 400 °C. [Figure 2] shows the background counts using the same un-irradaited sample as used for IR measurement above but with stimulations involving both heat and light. The optical stimulation was carried with samples being held at different temperatures, 150°C to 450°C in steps of 50°C, apart from the room temperature. The background can be seen to be around 100 counts per channel of 0.5 s at room temperature with 90% stimulation intensity using RisØ reader. However, the background recorded at elevated temperatures on the same sample could be seen to increase significantly with stimulation temperature. This can be seen to be extremely larger than IR contribution at these heating temperatures as shown in [Figure 1]. The background is found to increase/lift by a factor of 4 at 400 °C as compared to that recorded at 25 °C. The similar trend was seen in case of other sample also. OSL signal at elevated temperatures: The OSL signal has been observed to increase with stimulation temperature and this phenomenon is called as TA-OSL.[1] To study the dependence of OSL on the stimulation temperature, the CW-OSL signal was recorded, on Al2O3:C sample irradiated with 100 mGy dose, at stimulation temperatures of 250 to 350 oC, in steps of 50 oC as shown in [Figure 3]. For comparison, this measurement was also carried out at room temperature of 25 oC, as shown in inset of [Figure 3]. The OSL was seen to increase continuously with the increase in the stimulation temperature as shown in [Figure 3]. But, in order to evaluate the actual increase in TA-OSL signal, one must subtract the background signal from the total signal at that respective temperature. As described above, this background also increases significantly with the increase in the stimulation temperature in case of un-irradiated sample.{Figure 24}{Figure 25}{Figure 26}
Conclusion: The phenomenon of large background signal in an un-irradiated phosphor specimen with simultaneous optical and thermal stimulation is reported in this work. Studies involving TA-OSL should undertake this background emission into consideration for reliable results.
Keywords: Background signal, optical stimulation, TA-OSL
Reference
McKeever SW, Bøtter-Jensen L, Agersnap Larsen N, Duller GA. Temperature dependence of OSL decay curves experimental and theoretical aspects. Radiat Meas 1997;27:161-70.
Abstract - 41594: Assessment of effective radiation dose due to 222Rn in ground water to the population of Kodagu District, Karnataka State, India
S. N. Namitha, B. S. K. Lavanya, Mohammed Hidayath, M. S.Chandrashekara
Department of Studies in Physics, Manasagangotri, University of Mysore, Mysuru, Karnataka, India
E-mail: [email protected]
Radon, a decay product of radium is found in trace amount in natural water. Radon and its decay products in the atmosphere are the key contributors of human exposure to radiations from natural sources and pose a serious health risk.[1] Radon is identified to be the second largest cause of lung cancer. The ingestion of radon through drinking water pathway gives rise to an additional exposure dose to the stomach and other organs of the body. Airborne radon can be released from water during normal activities, such as showers, washing clothes and utensils, and so on. There is a strong correlation between radiation exposure and health hazards among the population in a given environment.[2],[3] In the present study, ground water samples were collected from various locations of Kodagu District, Karnataka State, India to measure the 222Rn activity concentration. Kodagu district occupies an area of 4102 km2 and comprises of rocks like granites, dykes, amphibolites and gneisses. The region receives an average rainfall of 2500 – 2800 mm per year. Ground water fulfils the major water demand of the population of study region. Therefore, it becomes necessary to estimate the concentrations of 222Rn in ground water and henceforth to estimate the radiation dose to the public. To analyse 222Rn activity concentration in ground water about 60 ml of sample was collected from each location of the study area during 2021 and 2022. 222Rn activity in collected water samples was measured by emanometry technique employing Smart Radon Monitor (SRM) which was calibrated regularly using a standard source. Air was bubbled through the water sample followed by transferring the radon gas into the Scintillation cell of the SRM. Activity concentration of 222Rn in water was calculated using the following relation,
[INLINE:1]
Where, D = Sample counts – Background counts, V = Volume of water (60 ml), E = Efficiency of the scintillation cell (74 %), t = Counting duration (s), λ = decay constant for radon (2.098 x 10-6 s-1), T = Counting delay after sampling (s). 222Rn concentration in ground water was found to be in the range of 0.22±0.39 to 8.61±5.73 Bql-1 with an average value of 1.75±2.66 Bql-1. The measured concentrations are found to be well below the recommended safety standards of 11.1 Bql-1 and 100 Bql-1 prescribed by USEPA and WHO respectively. Ingestion and inhalation doses due to 222Rn in ground water were calculated from the measured 222Rn concentration and are shown in [Table 1]. The total effective dose was found to vary from 0.77 to 29.84 μSvy-1.{Figure 27}{Table 12}
Keywords: Emanometry, ingestion, inhalation, radiation dose, radon
References
UNSCEAR. Annex B: Exposures from Natural Radiation Sources. United States: United Nations Scientific Committee on the Effects of Atomic Radiation; 2000.Mays CW, Rowland RE. Cancer risk from the lifetime intake of Ra and U isotopes. Health Phys 1985;48:635-47.BEIR. Health Risks from Exposure to Low Levels of Ionizing Radiation. National Academies Report in Brief; 2005. p. 7.
Abstract - 41597: Studies on 226Ra and 222Rn concentration in groundwater samples of Chamarajanagar District, Karnataka State, India
B. S. K. Lavanya, S. N. Namitha, Mohamed Hidayath, M. S. Chandrashekara
Department of studies in Physics, University of Mysore, Mysuru, Karnataka, India
E-mail: [email protected]
Exposure to natural radiation is an inevitable process to human being. Naturally occurring radionuclides have been present since the formation of the Earth. As ground water is in direct contact with these radionuclides, trace amounts of dissolved radioactive elements are present in it. When the ground water is used for drinking purposes, they cause serious health issues including cancer risks.[1] 226Ra is a naturally occurring radionuclide from the decay series of uranium and has a very longer half-life. It decays into 222Rn by emitting alpha particles that have enough energy to damage living cells. 226Ra is chemically similar to that of calcium and tends to accumulate in bones and teeth. It is not metabolised by the body; it only decays radiologically over time. 222Rn is a radioactive gas with a half-life of 3.82 days and decays by emitting alpha particle that pose inhalation and ingestion dose to the population. About half of the natural radiation is solely due to 222Rn and second largest contributor of lung cancer. 226Ra and 222Rn in water cause multiple health hazards.[2] The study area is Chamarajanagar district, lies in the southern tip of Karnataka State, India between north latitudes 11°40'58” and 12°6'32” and east longitudes 76°24'14” and 77°64'55“. The district consists of five taluks and the major rock types in the district are Granitic rocks and gneiss rock. People in the study area are mainly depending on ground water sources for drinking purposes. Therefore, distribution of radioactive elements in the groundwater and radiation dose to the public due to these radionuclides were studied. For 226Ra analysis emanometry method was employed. Water sample was pre-concentrated by co-precipitation and evaporation method. 70ml of solution was filled in a bubbler and kept undisturbed for 21 days. The air in the bubbler was transferred to the scintillation cell and counted for alpha activity using programmable counting system. For 222Rn measurements, water samples were collected in vials, up to the brim without air bubbles from sampling stations. The measurements are done within 4-5 hours of sampling using Smart Radon Monitor [SRM] based on the detection of alpha particles emitted from radon and its decay products in the scintillation cell. The measurements were done during 2021-2022 covering all the seasons. The concentration of 226Ra in ground water samples, measured at 12 regions of Chamarajanagar district covering all the taluks varied from 2.18 mBql-1 to 96.4mBql-1with an average value of 31.38mBql-1. These values lie well below the recommended limit of 1 Bql-1by WHO.[3] The 222Rn concentration varied from1.2Bql-1 to 11.65Bql-1with an average value of 4.25Bql-1,which is well below the recommended limit of 100 Bql-1by WHO. The total ingestion dose due to226Ra and 222Rn varied from 3.52 to 49.48 μSvy-1. The values suggest that there is no significant health risk to general public from these radionuclides in water. A strong correlation between 226Ra and 222Rn concentrations are observed with adjusted R-Square value of 0.9 indicate that they might be of same origin.
Keywords: Emanometry, ground water, ingestion dose, inhalation dose, SRM
References
UNSCEAR United Nations Scientific Committee on the Effects of Atomic Radiation. Sources and Effects of Ionizing Radiation. New York: United Nations; 2008Cothern CR. Radon, Radium, and Uranium in Drinking Water. CRC Press; 2014WHO World Health Organization. Guidelines for Drinking Water Quality-4th Edition, Radiological Aspects. World Health Organization; 2011. p. 203-17.
Abstract - 42116: Estimation of positional errors in direct measurement of 241Am in axillary lymph nodes
Lokpati Mishra, M. Y. Nadar, I. S. Singh, P. D. Sawant
Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
241Am is used as tracer for in-vivo assessment of Pu deposited in body.[1] In case of wound injury on hand with insoluble Pu/241Am, it shows tenuous retention at the wound site and then gets slowly transferred to axillary lymph nodes, blood and subsequently deposit in systemic organs (skeleton, liver etc).[2],[3] Hence, it is essential to precisely estimate the 241Am content in lymph nodes as it would interfere in 241Am measurements in Lungs (inhalation) and Liver (organ of deposition). To evaluate detection efficiency as well as to quantify the positional error in axillary lymph nodes measurements following experiments were performed using phoswich detector located inside totally shielded steel room. 241Am source was placed in the axillary lymph node location of the LLNL phantom and detector was placed above it. The estimated efficiency of system for activity deposited in the right and left lymph node are 2.3 E-02 cps Bq-1 and 2.1 E-02 cps Bq-1, respectively. Axillary lymphatic system in homo sapiens is a complex arrangement of several lymph nodes arranged in these three groups named apical, central and lateral in the shoulder region stretch from the arm pit towards the clavicle.[4] Due to difficulty in locating the axillary lymph nodes under arm pit, positional errors from reference lymph node position were estimated by sequentially moving the phantom with respect to detector in horizontal (± X) and vertical (± Y) directions [cf. [Figure 1]]. Similarly, error due to angular movement of the detector were estimated in clockwise and anticlockwise directions by rotating the phantom. Maximum observed deviations are given below [Table 1]. Maximum deviation of ~ 20 % was observed from the reference point when detector was moved away from the reference measurement position on phantom in +X and +Y direction. This is mainly because the photon falling on the detector active volume decreases as the detector moved away from the source. Additionally, the distance between detector surface and phantom surface changes due to non-uniformity in geometry of tissues near the lymph node. When the detector is moved towards the body of phantom (-X, -Y), deviation observed is significant (~50 %). This is mainly due to higher attenuation as well as scattering of the photons by the body tissue before reaching the detector surface. Attenuation is more in case of movement of detector in –X and –Y direction as detector position changes from lymph node to lungs. Thus, the observed deviation is dependent on attenuation of the photons within the body tissue as well as position of lymph nodes with respect to the detector. The developed methodology is extremely useful for assessment of Pu/ 241Am intakes due to wound injury and also in estimating interferences in lung monitoring due to radioactivity present in lymph nodes.{Figure 28}{Table 13}
Keywords: 241Am, axillary lymph node, phoswich detector, positional error
References
Mishra L, Singh IS, Patni HK, Rao DD. Comparing lungs, liver and knee measurement geometries at various times post inhalation of 239Pu and 241Am. Radiat Prot Dosimetry 2018;181:168-77.Singh IS, Mishra L, Yadav JR, Nadar MY, Rao DD, Pradeepkumar KS. Applying a low energy HPGe detector gamma ray spectrometric technique for the evaluation of Pu/Am ratio in biological samples. Appl Radiat Isot 2015;104:49-54.Nadar MY, Patni HK, Akar DK, Mishra L, Singh IS, Rao DD, et al. Monte Carlo simulation of embedded 241Am activity in injured palm. Radiat Prot Dosimetry 2013;154:148-56.Zeman R, Lobaugh M, Spitz H, Glover S, Hickman D. A calibration phantom for direct, in vivo measurement of 241Am in the axillary lymph nodes. Health Phys 2009;97:219-27.
Abstract - 42140: Internal dose calculation code in line with the ICRP 2007 recommendations
K. Manabe, F. Takahashi
Nuclear Safety Research Center, Japan Atomic Energy Agency, Tokai-mura, Ibaraki, Japan
E-mail: [email protected]
The Japan Atomic Energy Agency has been developing a code for internal dose calculation in line with the 2007 Recommendations of the International Commission on Radiological Protection (ICRP). The code named as Internal Dose Calculation Code (IDCC) has two major functions: calculation of dose coefficients and estimation of intakes of radionuclides from monitoring data. The function calculating dose coefficients is developed mainly to revise the Japanese regulatory standards for internal exposure.[1] The existing standards are established as derived air concentrations for workers and concentrations of exhausted air and drain water from facilities for the public, which are determined from the effective dose coefficients based on the 1990 Recommendations of the ICRP for each radionuclide and its chemical form. Although the ICRP does not provide dose coefficients for radionuclides with half-lives of less than 10 minutes, the regulatory standards include values for those radionuclides. Therefore, we need a code to calculate dose coefficients in line with the ICRP 2007 Recommendations. The developed function was verified by confirming that the generated dose coefficients coincide with the values recorded in OIR (Occupational Intakes of Radionuclides) Data Viewer of the ICRP. A dosimetry tool in conformity with the ICRP 2007 Recommendations is necessary for radiation safety management and for retrospective assessment in an intake incident. Then, we have also developed a function estimating intakes of radionuclides by using individual monitoring data obtained from in-vivo measurement and bioassay.[2] This function is applicable to the conditions of a single acute intake, multiple acute intakes, and chronic intakes. The maximum likelihood method is applied to estimate intakes in fitting the time course of the activities in the body and organs and/or excretion rate generated by the code to monitoring data. The function is verified by confirming that the estimated intakes coincide with the literature values for case studies of IMBA code and the IDEAS guidelines. The function estimating intakes will be used not only by researchers and technical staffs of radiation protection but also local government staffs concerning nuclear emergency preparedness. Therefore, usability is important. Then, we developed a graphical user interface (GUI) to integrate the two functions and to operate them. [Figure 1] shows samples of the windows of the GUI of the IDCC. Users can select calculation modes from calculating dose coefficients or estimating intakes, and can set calculation conditions instinctively by using pull-down menus. The GUI also enables to edit data of biokinetic models and specific absorbed fractions (SAF). The data editing capability is useful for individual retrospective dose assessment. In addition, installers for Windows, Mac and Linux PCs were developed for ease of use. Development of the arithmetical operation part of IDCC has been completed. Now, we are working to incorporate the biokinetic models of ICRP Publ. 151. In the future, we plan to implement data for members of the public and SAF data for the average adult Japanese[3] to IDCC. In the presentation, we will introduce the overview of the IDCC and the distribution of the IDCC. This work was funded by the Nuclear Regulation Authority of Japan (JPJ007057).{Figure 29}
Keywords: Calculation code, dose coefficient, intake estimation, internal exposure
References
Manabe K, Sato K, Takahashi F. Bio Conf 2019a;14:03011.Manabe K, Sato K, Takahashi F. 5th International Symposium on the System of Radiological Protection; 2019b.Manabe K, Sato K, Takahashi FJ. Nucl Sci Technol 2022;59:656-64.
Abstract - 42156: Standardization of a liquid scintillation counting based methodology for quantification of 125I in urine samples
Sonali P. D. Bhade1, Rajesh Sankhla1,2
1Radiation Safety Systems Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
125I is usually used in clinical medicine for radio-labelling and tracer studies. It decays by Electron Capture (EC) with a half-life of 60.1d. Once Iodine reaches the blood, 30% is accumulated in thyroid while remaining 70% gets excreted directly via urine. Workers handling radioiodine, are monitored routinely for internal contamination. Generally, in-vivo monitoring of 125I is carried out using thin NaI(Tl) based Thyroid Monitoring System. The objective of the study was to develop Liquid scintillation counting (LSC) based methodology for quantification of 125I in urine samples, as an alternate technique to thyroid monitoring. 125I decays by EC followed by emission of auger electrons [2.3-4.8keV (Auger-L) and 21.8-31.7keV (Auger-K)], which deposit their energy in the scintillation cocktail. The excited solvent molecules pass on their excitation energy to flour molecules which upon de-excitation emit light photons that are detected by PMTs. The intensity of signal thus generated is proportional to the energy of auger electrons. Urine is a complex bodily fluid with regard to its constituents. The components of the urine have the potential to pose colour as well as chemical quenching in LSC measurements. Hence, quenching effect on the quantification of 125I activity in urine samples need to be studied crucially. Firstly, 125I source was standardized using CIEMAT/NIST efficiency tracing technique.[1] In the present study, Quantulus 1220 LSC was used for 125I measurements. Urine samples, collected from the unexposed individuals, were spiked with 125I standard (125I activity: 100Bq) and categorized in 3 different sets. 1st set consisted of spiked urine samples, quenched with brown dye (colour quench studies) while in the 2nd set, Nitromethane (chemical quench studies) was added to the spiked samples and in the 3rd set both colour and chemical quenchers were added in incremental amount in spiked urine samples to simulate several quench levels. First two sets were employed for calibration of the LSC system, while urine samples in the 3rd set were analyzed later to validate the standardized procedure for quantification of 125I. These samples were mixed with Optiphase Hisafe-3 scintillation cocktail in HDPE vials in 8:12 proportion. Quantulus LSC uses Spectral Quench Parameter (SQP(E)) as Quench Indicating Parameter (QIP). Each spiked urine sample, in the above sets, was counted in the set ROI (40-360channels, corresponds to 1-31.7 keV) for a counting time of 30min. For the urine samples in the first two sets, Counting Efficiency (CE) was calculated and plotted against respective, SQP(E) parameter [Figure 1]. This calibration plot was employed further to arrive at the CE values subsequent to level of quenching in the spiked urine samples. Analysis of urine samples in the 3rd set, using the above calibration plot indicated that for samples with QIP in the range of 600 < SQP(E) <710, results obtained using both colour and chemical plots were comparable (±10%). However, for higher degrees of quench (SQP(E)<600), significant deviation (>20%) was observed using colour quench plot. Colour had a larger influence on the CE values, as compared to chemical components, for the same degrees of quench [Figure 2] and hence, higher deviation was observed using color calibration plot. The present study indicated that chemical calibration plot provides precise results irrespective of varied quench levels in urine samples. Minimum Detectable Activity (MDA), achieved for 125I, by this method is 3 BqL-1 (counting time: 500min). This technique does not require any chemical separation and is sensitive to detect CED < 1mSv within 3 days of any incidental exposures.{Figure 30}{Figure 31}
Keywords: 125I, Liquid Scintillation Counting, quenching, spiked urine samples
Reference
Malonda G, Garcia-Torano E. Int J Appl Radiat Isot 1982;33:249-53.
Abstract - 42214: Validation of quick scan whole body monitor using FLUKA detector modelling
C. S. Charubala1, S. L. Asanali1, V. Santhanakrishnan1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
The whole body counter uses NaI(Tl) or HPGe detectors for the estimation of activity due to high energy photons emitted by fission products. The efficiency calibration of the whole body counter is carried out using BOMAB phantom. Numerical calibration methods may be employed to evaluate the uncertainties in physical calibration of these systems. In this study, FLUKA numerical model of Quick Scan Whole Body Monitor (QSWBM) was validated by experimental source response measurements to employ the model in numerical calibration using voxel phantoms. QSWBM consists of two NaI(Tl) detectors of dimension 406 mm x 127 mm x 76 mm housed in shielded structure for fast scanning the radiation workers to detect high energy photon emitters as quick as in 60 seconds in standing linear geometry.[1] Experimentally [Figure 1]a, QSWBM response was evaluated using two disc sources, 137Cs (661.7 keV) and 60Co (1173 keV and 1332 keV) of activities 38 kBq and 66 kBq as on 01/02/2015 respectively. For FLUKA[2],[3],[4] simulations [Figure 1]b, the BEAM card was used to simulate 661.7 keV, 1173 keV and 1332 keV energies. DETECT card was used to score energy deposition in both the detector regions with energy binning same as that in experiment (5keV/channel). To compare the simulation with experimental results, Gaussian smearing was performed using a user routine in which A, B and C coefficients of experimentally derived Full Width at Half Maximum (FWHM) function was incorporated. Counting efficiency (CE) from experiment and simulations were estimated from the sum of net counts recorded in two detectors. The results were compared to validate the FLUKA model of QSWBM.{Figure 32}
The CEs estimated from the experiment and simulations are summarized in [Table 1]. It can be seen that the percentage deviation for all energies are less than ±4% which is a reasonably good agreement for numerical model validation. The statistical uncertainties for all experimental measurements were <1% and relative error for simulations were <3%. A comparison of detector spectra from experiment and simulations is shown in [Figure 2]. The simulated spectra using FLUKA detector models of this study were found to have good agreement with the corresponding experiment spectra except a small channel shift for the higher energies of 60Co.{Figure 33}
Keywords: Calibration, FLUKA, in vivo monitoring, whole body counting
References
Sankhla R, Singh I, Rao DD, Pradeepkumar KS. Development of Quick Scan Whole Body Monitor for in-vivo Monitoring of Radiation Workers and General Public. BARC Newsletter; 2015.Ahdida C, Bozzato D, Calzolari Widorski M. New capabilities of the FLUKA multi-purpose code. Front Phys 2022;9.Battistoni G, Boehlen T, Cerutti F, Chin PW, Esposito Smirnov G. Overview of the FLUKA code. Ann Nucl Energy 2015;82.Vlachoudis V. Flair: A Powerful but user Friendly Graphical Interface for FLUKA; 2009. p. 2.
Abstract - 42232: In-vivo monitoring for a novel radionuclide
H. K. Patni, D. K. Akar, V. P. Ghare, P. D. Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai,
Maharashtra, India
E-mail: [email protected]
In-vivo/direct measurement of radionuclides in person's body is carried out to estimate dose due to internal contamination. Penetrating radiations like gamma/x-rays emitted from person's body are detected using photon spectrometers e.g. NaI (Tl), HPGe etc. in shielding. Counting efficiency (CE) of these systems are estimated using phantoms viz. BOMAB (with 133Ba, 137Cs & 60Co sources) or realistic thorax phantoms (with 241Am, U, 239Pu sources). The count rate from person's body with CE of the identified radionuclide is used to estimate the radioactivity present in person's body. This measured radioactivity is used to arrive at intake and committed effective dose (CED) using retained fraction (RF) and dose coefficients, respectively. If a novel radionuclide is identified in some person, then its dose estimation will require quick estimation of CE and RFs. Monte Carlo simulations are a robust method to compute CE and the RFs for intake estimation can be derived using biokinetic model solution. Recently, BARC has developed 106Ru based novel brachytherapy sources[1] for eye cancer treatment. The potential for internal contamination of 106Ru during source preparation process exists. In this paper, in-vivo monitoring of 106Ru is shown as an approach to be followed during in vivo monitoring of any novel radionuclide. 106Ru (T1/2=371.5 days) undergoes pure beta decay into 106Rh (T1/2=30.1 sec) which further undergoes beta decay and emits 511.9 (20.4%) and 621.9 keV (9.9%) photons. These photons can be used for identification and quantification of 106Ru by in-vivo monitoring. Although, 511.9 keV photons have higher yield, its counts suffer interference from 511 keV annihilation photons and hence, is not preferred for estimation. Monte Carlo simulations in FLUKA code was carried out for partially shielded system incorporating 102 mm dia. x 76 mm thick NaI (Tl) detector. ICRP reference male voxel phantom[2] (73 kg/176 cm) was modified into Indian reference worker dimensions (66 kg/167 cm). Height ratio was used to modify slice thickness and this in conjunction with weight ratio was used to modify voxel in-plane resolution. Uniform distribution of source points in the modified phantom and isotropic emission of 621.9 keV photons from these points were simulated using random sampling technique. 34 equidistance static positions (from top of head at 5 cm, 10 cm, ..., 170 cm) were assumed to simulate whole body scanning process (CE already validated for 137Cs, 133Ba and 60Co in Akar, 2013).[3] [Figure 1] shows geometry of the monitoring system with phantom, as used in FLUKA simulations. Detect card was used to arrive at the pulse height spectra in the detector due to emissions in the phantom. Systemic model for 106Ru was taken from ICRP-137. After entry into blood ruthenium deposits in liver, bone, urinary bladder and GI tract. Revised HRTM and HATM parameters were used to build the entire biokinetic model of 106Ru. Since, 106Rh is extremely short lived it is safe to assume that it will follow the biokinetics similar to Ru. The model was solved for inhalation of Type M 106Ru using in-house developed and validated octave program to estimate amount of RFs. Dose coefficient for inhalation of type M 106Ru is 1.3E-08 Sv/Bq (ICRP 137). The CE estimated using simulations is 4.96 cpm/kBq with 0.5% relative error for 621.9 keV photons of 106Rh. Using person background and CE, the MDA was computed as 1.1 kBq for 20 min of monitoring. The whole body RF for annual monitoring frequency was estimated as 1.6E-2 Bq for inhalation of 1 Bq type M 106Ru. Ratio of dose coefficient and RF gives a quantity dose per content which is 8.2E-07 Sv/Bq in this case. The corresponding CED is 0.89 mSv. The developed method for estimation of CE can be used to estimate radioactivity in person's body for quantification of radionuclide. The method for solving biokinetic model to estimate RFs can be used when they are not available in standard tables or behaviour inside body is differing much from reference values. By using these factors and dose coefficients CED can be computed.{Figure 34}{Table 14}
Keywords: Biokinetic model, In vivo monitoring, Monte Carlo simulations, voxel phantom
References
Sinharoy P, et al. Sep Sci Technol 2021;56:1450-6.ICRP. Adult Reference Computational Phantoms. London: SAGE; 2010.Akar DK, et al. Radiat Prot Dosimetry 2013;155:292-9.
Abstract - 42234: Estimation of retained fractions of 131I during pregnancy
Vandana P. Ghare, H. K. Patni, D. K. Akar, P. D. Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
131I is an important radionuclide in nuclear industry. It is widely used in medical industry for both diagnostic as well as therapeutic applications. Due to its volatile nature, there are chances of internal exposure while handling of 131I. It has also been an important radionuclide released during the past nuclear accidents. Hence, additional precautions are needed for pregnant woman to reduce the fetal dose. ICRP-56 has given systemic model of iodine for various age groups and adults, but not for pregnant women. ICRP-88 (2001) has provided biokinetic model and dose coefficients for pregnant women and fetus for ingestion pathway at different times during pregnancy. Using this model, Berkovski et al.[2] discussed about dosimetry for embryo and fetus following ingestion of 131I. Although, inhalation is the major pathway for radiation workers, in case of radiation accident, ingestion pathway becomes the major route of intake of 131I. In case of internal contamination monitoring of a pregnant woman with 131I, retained fractions are required for intake and dose estimation. These retained fractions are presently not available in the literature. Hence a study was taken up to estimate thyroid retention fractions of 131I in pregnant woman due to inhalation and ingestion of unit activity at various time intervals during pregnancy. ICRP 88 biokinetic model of iodine for pregnant woman is more complex than age dependent ICRP 56 model. It takes into account bidirectional transfer of iodine from mother to fetus through placenta, retention of iodine in placenta and uptake of iodine by amniotic fluid & fetus and other metabolic changes during pregnancy. Model gives information about absorption rates of radioiodine by mother and fetal thyroid. Time independent and time dependent (time from conception) transfer rates are given in the model. The model given in ICRP 88 along with ICRP 66 respiratory tract model was combined to solve for inhalation (Vapour Class SR1) of 131I. For ingestion pathway, model described in ICRP 88 was used. These models were incorporated in octave program developed for solving biokinetic model of 131I at various time intervals viz. 0th week (at conception), 3, 8, 12, 16, 24, 32, 36 and 38 weeks after the onset of pregnancy. Thyroid retained fractions were estimated in pregnant woman at these time intervals of pregnancy for 1 to 15 days after the intake. Thyroid uptake by fetal thyroid was also estimated at different times during pregnancy. Graphs for thyroid retained fractions, in case of inhalation and ingestion of 131I by adult worker and pregnant woman at various times of intake during pregnancy are shown in [Figure 1]. The graphs show thyroid retained fractions from 1 to 15 days for intake at 0, 3, and 38 weeks. It is observed that for ingestion route, if the intake happens at 0th week, then mother thyroid retained fraction on 1st day post intake is 0.25 and in case of intake in 3rd week to 38th week, variation in thyroid retained fraction for pregnant woman is not significant (0.38 to 0.35, respectively). In case of inhalation route, thyroid retained fraction for pregnant woman on 1st day post intake is 0.23 at 0th week and varies from 0.34 to 0.31 at 3rd week to 38th week. It is observed that in initial weeks of pregnancy (upto 12th week), there is no uptake by fetal thyroid. For inhalation route, fetal thyroid retained fraction on 1st day post intake at 12, 16, 24, 32, 36 and 38 week are 0.0002, 0.0016, 0.0089, 0.031, 0.052, 0.069 respectively. The retained fractions estimated in the present study are useful for assessment of intake and internal dose to pregnant women (radiation worker as well as member of public). The fetal retained fractions will also help in estimating the dose received by the fetus in case of internal contamination of the mother or during her treatment with 131I.{Figure 35}
Keywords: 131I, fetus, pregnancy, retained fraction
References
ICRP. Doses to the Embryo and Fetus from Intakes of Radionuclides by the Mother. ICRP Publication 88. Annals of the ICRP; 2001.Berkovski V, et al. Radiat Prot Dosimetry 2003;105:265-8.ICRP. Human Respiratory Tract Model for Radiological Protection, ICRP Publication 66. Annals of the ICRP; 1994.
Abstract - 42238: An alternative rapid method for 131I estimation in bioassay samples
Seema Chaudhary, Sonal Wankhede, Suja A. Kumar, Rahul Roy, Amar Dutt Pant1, Prathibha Pradosh, Pramilla D. Sawant
1Radiation Safety Systems Division, Bhabha Atomic Research Centre, 2Environmental Monitoring and Assessment Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Radioiodine (131I) is a major fission product, comprising ~ 3% (by wt) of the total fission products. In case of any radiation accident (nuclear power plant and facility handling radioiodine), 131I being volatile gets easily released into the environment. This may result in thyroid uptake of radioiodine by workers and / or members of the public. The most probable route of 131I intake for workers is inhalation and for the population, it's by ingestion of contaminated foodstuff. Once iodine enters the body through whichever route, it gets absorbed into the blood and then a fraction (0.3) gets accumulated in the thyroid gland and remaining (0.7) is excreted directly in urine.[1] 131I emits a photon of 0.36 MeV following a beta decay. This photon energy is thus used for identification as well as quantification of 131I in urine using gamma spectrometry technique. The conventional method used at Bioassay Lab, Trombay for estimating radioiodine in urine involve the following steps: 10 mg iodide carrier and 200 mg silver chloride are added to acidified urine and stirred for 2 h. Iodide ion in the solution replaces chloride ion in AgCl and gets precipitated as silver iodide. This precipitate is then dissolved in 2% KCN, transferred to a γ-counting vial and counted using a gamma spectrometer. Organic fraction of iodine in urine is estimated by converting it to iodide using acidified KMnO4 and then the procedure as mentioned above is followed. As the conventional procedure for estimation of 131I in bioassay samples is time consuming and laborious, a need was felt to standardize an alternative rapid method. For this purpose, 131I standard solution procured from RPhD (2 MBq, Consignment no. 4172) was diluted to a known volume and standardized by gamma spectrometry. 3 L urine samples was collected from a healthy individual (member of the public) and divided into aliquots of 20, 100 and 200 mL. These aliquots were spiked with 131I activity (0.6 kBq) and the spiked activity was standardized for these three geometries using HPGe detector (30% relative efficiency). 20 mL urine sample was taken up in glass scintillation vial and 100 mL and 200 mL were placed in a polythene container geometry. Efficiency of 131I in HPGe detector was observed to be 3.7% and 1.42% for 20 mL glass vial and 200 mL polythene container, respectively. The background of the system was determined by using same urine samples aliquot but without spiking any 131I activity. Minimum Detectable Activity (MDA) for 200 mL blank urine sample is 9.0 Bq/d for 1h counting time. The developed method is sensitive to detect Committed Effective Dose (CED < 1 mSv) even 20 days after the inhalation of Type F compounds of 131I. In case of spot samples with insufficient sample volume (< 100 mL), 20 mL glass vial geometry is preferrable but samples need to be counted for 5 h. The root mean squared error for this method is 0.02 which is lower than the performance criteria (< 0.25) recommended by ANSI N13.30 for testing laboratories. For further validation of the method developed, analyst involved in method development participated in an interlaboratory intercomparison exercise. As a part of the exercise, total eight urine samples (volume 100 mL and 200 mL) were spiked with 131I activity and distributed to four analysts from three different laboratories. Analyst no 1-3: Directly counted the sealed 200 mL sample in 30% HPGe. For 100 mL sample, a 20 mL aliquot was taken up in glass vial and counted using HPGe detector. 20 mL aliquots were also counted using CsI(Tl) based portable thyroid monitoring system (PTM). Analyst no. 4, counted the samples directly in well type NaI(Tl) and NaI(Tl) based Shadow Shield Whole Body Monitor (SSWBM). Results of the intercomparison exercises are given in [Table 1]a and [Table 1]b. The relative bias (‒1.0 %) and the relative precision (2.4 %) are within the acceptable criterion recommended by ANSI N13.30 for radio-bioassay (−25% to +50% for relative bias and 40% for relative precision). Participation in the intercomparison exercise has helped in validating the methodology developed for rapid estimation of 131I in urine samples.{Table 15}
Keywords: Gamma spectrometry, iodine, method validation
Reference
International Commission on Radiological Protection. ICRP Publication 137; 2017.
Abstract - 42239: Intercomparison exercise for estimating 90Sr in emergency bioassay samples
Sonal Wankhede, Seema Chaudhary, Soumitra Panda, Rupali Dubla and Pramilla D. Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
In case of an emergency scenario, large number of bioassay samples need to be analysed and therefore a rapid radiochemical separation method would be required for analysis and reporting of results. A rapid procedure using solid extraction chromatography was established for estimation of 90Sr in urine at Bioassay Lab, Trombay in 2014.[1] In the present study, further reduction in analysis time was achieved by standardizing solid extraction chromatography method using vacuum box system. This method uses 2mL prepacked Sr Spec column and radiochemical separation is done by employing vacuum. Average radiochemical recovery obtained by this method is ~ 92%. Both these rapid methods were validated by participating in inter-laboratory intercomparison exercise for emergency bioassay. The intercomparison exercise was conducted to check the response capabilities in assaying 90Sr in urine, focusing on the procedures and methods/techniques used and the time required in submitting results by the participating laboratories. Intercomparison sample was prepared by pooling urine samples collected from three healthy individuals not occupationally exposed to any radionuclides. The pooled sample was further divided into 100 mL aliquots and spiked with 90Sr standard (37 kBq/mL, BRIT-No. 5279). These spiked samples along with a blank sample were distributed to 4 analysts from 3 different laboratories after adding 1% HCl. Methodologies adopted by the analysts for estimating 90Sr in these samples were as follows: (1) Direct Liquid Scintillation Counting (LSC) after mixing an aliquot of the samples with Ultima Gold uLLT scintillation cocktail. (2) Solid Extraction Chromatography (SEC) technique using 1 g of Sr spec resin in glass column and radiochemical separation performed under gravity followed by LSC. (3) Solid Extraction Chromatography (SEC) technique using 2 mL pre packed Sr column and radiochemical separation was performed by employing vacuum followed by LSC. Liquid Scintillation Spectrometry instrument used by the analyst were 300 SL TDCR Liquid Scintillation Counter (make Hidex), Ultra low level Quantulus 1220 (make M/s Perkin Elmer) and Ultra low level Quantulus GCT (make M/s Perkin Elmer). Results of the 90Sr analysis along with the spiked values are given in [Table 1]. Method accuracy was assessed using relative bias (Br) as defined by ANSI N13.30. It mentions the acceptance criterion for radio-bioassay as −25% to +50% for Br and 40% for relative precision (SB), respectively. In case of samples E6b & 30-A, relative bias for individual samples (Bri) was observed to be outside the range given by ANSI. Hence, the analysts were informed to make appropriate corrections in analysis protocols to reduce or eliminate bias. Overall Br and SB observed was −4.5% and 5.0 %, respectively. The root mean squared error (RMSE) observed in the present exercise was 0.07 which is less than the value recommended by ANSI 13.30 (< 0.25). The response time for reporting the results was <1 working day for all the three Laboratories.{Table 16}
Keywords: Bioassay, intercomparison, strontium, vacuum box
References
Wankhede S, et al. Radiat Prot Environ 2014;37:95-100.ANSI 13.30 Performance Criteria for Radiobioassay; 1996.
Abstract - 42244: Gamma spectrometry of urine sample instead of whole body during COVID-19
D. K. Akar, H. K. Patni, V. P. Ghare, R. Sankhla, I. S. Singh, L. Mishra, P. D. Sawant, P. Chaudhury
Radiation Safety Systems Division, Bhabha
Atomic Research Centre, Mumbai,
Maharashtra, India
E-mail: [email protected]
Whole body monitoring (WBM) is an important technique for internal contamination monitoring for radionuclides which emit penetrating radiation. It helps to ensure regulatory requirements are fulfilled during handling of radioactivity. WBM systems have low background due to partial or full shielded enclosures with a provision to position suspected contaminated person in counting geometry near the photon detector. For WBM, spectrum acquired from person's body is used for radionuclide identification followed by activity, intake and dose estimation. During Covid-19 outbreak, continuation of WBM for essential activities of BARC was required. In such scenario, WBM of the workers following social distancing norms became challenging due to following issues, i) unknown status of infection in involved persons, ii) the positioning of workers in WBM system without close contact between worker and operator, iii) use of protective gear during monitoring in enclosed system for the workers, iv) time gap between subsequent monitoring for proper air changes and frequent sanitization of the shielded systems. International bodies like EURADOS, IAEA[1] recommended internal monitoring only for cases exceeding committed effective dose (CED) 1 mSv and use of methods where direct contact of persons is minimal. ICRP has recommended[2] gamma ray spectrometry of urine samples for monitoring of 131I, 137Cs and 60Co besides WBM technique. A comparison of these two techniques was carried out by Carbaugh et al., 2015.[3] In present study, setup and factors required for adopting gamma ray spectrometry of urine samples for monitoring of 131I, 137Cs and 60Co, are discussed. Shadow shielded WBM [Figure 1] was used for monitoring of urine samples. Urine samples were monitored in PVC made cylindrical bottles (dia 9 cm * height 21 cm) kept 1 cm away from the detector face. A 10.2 cm dia * 7.6 cm thick NaI(Tl) detector is housed in this WBM such that any ambient photons directed towards detector will be attenuated by 15 cm mild steel and 3 mm Pb. For a photon to reach detector after Compton scattering, required scattering angle is more than 90° (final photon energy < 511 keV). The system's counting efficiency (CE) was experimentally estimated for 1 litre aqueous solution with known activity of 137Cs and 133Ba (131I simulant) in the bottle. Monte Carlo simulations were also carried out in FLUKA code where experimental setup was replicated to compute and validate CEs. The simulations were performed for 60Co to estimate its CE. Uncontaminated urine sample and the CEs were used to estimate the background and MDA of the system for these radionuclides for 30 minutes of counting time. Urine sample collection duration was set to be overnight. A factor of 2 is used to convert measured activity in the sample into activity excretion per day. The measured values with ICRP predicted excretion rates and dose coefficients can be used to compute intake and CED. The estimated CE values are 445, 763 and 701 cpmkBq-1 and the corresponding MDAs are9.3, 10.3and 4.7 Bq for 137Cs, 131I and 60Co (for both peaks), respectively. The minimum detectable dose (MDD) for 360 days monitoring interval for 137Cs and 60Co are 0.18 and 1.08 mSv respectively. For WBM these values are 9 and 30 μSv respectively. For 131I MDD was 9 μSv for monitoring 2 days after intake with 2 μSv being corresponding value for thyroid monitoring. This method was applied to overcome the limitations posed by the pandemic situation and increased monitoring throughput, that helped in fulfilling the regulatory requirements. Present study shows that, counting of urine samples can be utilized as an alternative to WBM of a person, although the MDD is higher. The developed method can be further optimized using a gamma spectrometry system through appropriate counting time, geometry and shielding.{Figure 36}
Keywords: Gamma spectrometry of urine, whole body monitoring
References
Vanhavere F, et al. EURADOS Recommendations to Deal with the COVID-19 Pandemic. Available from: www.iaea.org.ICRP. Annals of ICRP. ICRP Publication 78; 1997. p. 27.Carbaugh EH, et al. Health Phys 2015;109 Supp 2:S141-7.
Abstract - 42246: Comparison of counting efficiencies of quick scan whole body monitor using BOMAB and Voxel phantoms
Rahul Roy1, Rajesh Sankhla1,2, P. K. Singh1, J. R. Yadav1, P. D. Sawant1
1Radiation Safety Systems Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Whole body monitoring of radiation workers for internal radioactive contamination is an essential and regulatory requirement. In-vivo monitoring for high energy photon emitters (HEPs) (E > 200 keV) is carried out using partially shielded systems. A high throughput standing linear geometry Quick Scan Whole Body Monitor (QS-WBM) is commissioned at H. P. Lab., Tarapur. This system has been experimentally calibrated using anthropometric BARC reference BOttle Mannequin ABsorption (BOMAB) phantom which consists of ten PVC containers representing different body parts, when filled with water and assembled, it resembles the shape of human body. International Commission on Radiological Protection (ICRP) has provided reference voxel phantoms[1] which are more realistic and anthropomorphic in nature. In this study, theoretical counting efficiencies (CEs) for QS-WBM are estimated using BOMAB and Indian sized voxel (IND-VXL) phantoms, both having BARC reference dimensions (H: 168 cm, W: 67 kg)[2] and results are compared. The QS-WBM consists of a three-sided shielded cavity made up of 50 mm Pb and 6 mm SS plates, having outer dimensions of 2170 mm (H) × 1225 mm (W) × 625 mm (D). Two large size NaI(Tl) detectors each of dimensions- 406 mm (L) × 127 mm (W) × 76 mm (D) are placed one above another in a linear array inside detector mounting wall in front of the subject's standing platform.[3] The ICRP voxel (ICRP-VXL) phantoms are based on the CT scan or MRI data of real persons segmented into voxels and assigned with tissue equivalent material that represents human anatomy more accurately. ICRP-VXL phantom has height of 176 cm and 73 kg of weight. The IND-VXL phantom is constructed by reducing the in-plane resolution& height of the ICRP-VXL phantom as presented in [Table 1]. The QS-WBM geometry is incorporated in FLUKA along with BOMAB and IND-VXL phantoms separately as shown in [Figure 1]. In BOMAB phantom, activity is distributed in the axial cavities whereas in voxel phantom activity is distributed uniformly in whole body organs for 137Cs, 60Co and 40K. For sampling 131I in thyroid of BOMAB phantom, IAEA acrylic phantom is used. For 60Co, area of both peaks together is taken. Random point sampling is carried out by writing SOURCE ROUTINE in Fortran. Monte Carlo simulations are performed using 107 primaries (5 cycles) and energy deposition in detectors is scored using DETECT card of FLUKA code. The percentage relative errors in all simulations performed are less than unity. From [Table 2] it can be seen that, CEs for BOMAB phantom are less compared to IND-VXL phantom and varies up to ∼15%. It is because voxel phantoms are anthropomorphic in nature which makes some of its organs relatively closer to detector as compared to BOMAB phantom of same dimensions. Therefore, when CEs of QS-WBM for BOMAB phantom are used, whole-body content will be over-estimated by a factor of ∼1.1. These results will be useful in determining the effect of the realistic anatomical and compositional details of voxel phantoms on the CEs of the system.{Table 17}{Figure 37}{Table 18}
Keywords: Computational phantoms, counting efficiency, FLUKA, QS-WBM
References
ICRP. ICRP Publication 110 Adult Reference Computational Phantoms. Pergamon Press; 2009.Akar DK, et al. Comparison in Detection Efficiency of WBM Using BOMAB and Voxel Phantoms. IARPIC; 2018.Sankhla R, et al. Development of QS-WBM for Radiation Workers and General Public. BARC Newsletter; 2015.
Abstract - 42247: Estimation of dose coefficients for 131I using paediatric reference computational phantoms
P. K. Singh, H. K. Patni, Rahul Roy, D. K. Akar, P. D. Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Radiation dose from 131I to general public & occupational workers, in case of radiological and nuclear emergency, is a major concern in health physics. Accurate assessment of internal dose due to radio iodine therapy is also required for better treatment planning. Current ICRP dose coefficients (DCs) for general public are based on age specific Medical Internal Radiation Dose (MIRD) phantoms and ICRP-56 biokinetic models for Iodine. Recently, ICRP-143[1] has released 10 paediatric reference computational (PRC) phantoms representing male & female of different age groups (newborn, 1 Y, 5 Y, 10 Y & 15 Y). Age dependent systemic model for Iodine has been updated in 2017.[2] DCs based on PRC phantoms and revised iodine models are not yet available. In this study, 131I DCs for PRC phantoms are estimated for inhalation (1 μ AMAD) & ingestion cases using latest age-specific biokinetic parameters. Estimation of DCs involves two steps: (1) Estimation of number of disintegrations (US) occurring in all source regions. For this, ICRP-66 respiratory tract, latest alimentary tract (ICRP-100) & systemic[2] parameters are coupled to create age-specific biokinetic models. These age groups are same as the age groups of PRC phantoms. An octave program is written to solve inhalation & ingestion biokinetic models which yields US in each case. (2) Calculations of S-values (absorbed dose delivered to the target tissue per unit disintegration of the radionuclide in source regions, (Gy. (Bq s)-1). For this following scheme was used: First, using a Fortran code supplementary data files given in ICRP-143 are converted into PRC voxel files readable by FLUKA code.[3] Tissue equivalent media and their elemental composition are also assigned in various VOXEL ID numbers of all the 10 phantoms separately using another Fortran code. These voxel files are used as part of separate input files for MC simulations. For each phantom, Monte Carlo simulations are performed (5 spawns of 107 primaries) to compute the energy deposited in all target voxel regions, due to transformation in possible source voxels regions (as per the biokinetic model). The photons & electrons emissions of 131I are invoked by using HIPROPert & RADDECAY cards of FLUKA. Output from USRBIN card (provides absorbed fraction (AF) in target voxels) & target organ masses from ICRP-143 are used to compute various S-values for each phantom. These Us & S-values are used to estimate age-specific tissue equivalent doses (HT) to all 27 target tissues. Finally, Sex-averaged HT values & tissue weighting factors are used to calculate age-specific DCs for both routes of the intake for PRC phantoms. Relative errors in AF for PRC phantoms are 1% for most target voxels in simulations. From [Figure 1], it can be seen that, DCs decreases with age of phantom (about 86 & 83% for inhalation & ingestion cases respectively). This is due to the following reason: masses of target organs increase with age & S-values varies reciprocally with target masses. Therefore, S-values decrease with phantom age (for e.g., S(thyroid-thyroid) decrease from 2.29E-08 to 2.65E-09 for new born to 15 Y). Due to decrement in S-values, HT and hence DCs also decrease with age. Further, ingestion DCs are about 63% higher than DCs for inhalation as Us in case of ingestion are higher than inhalation. This is because for ingestion, blood uptake and hence absorption to thyroid, is relatively higher than inhalation. Estimated DCs can be used for radiation dose calculation in case of radiological and nuclear emergency. These values can also be used for dose calculation during radio iodine therapy in patients of different age groups.{Figure 38}
Keywords: 131I dose coefficients, Monte Carlo simulations, paediatric voxel phantoms
References
ICRP. ICRP Publication 143 Paediatric Computational Reference Phantoms. Annals ICRP; 2020. p. 49.Leggett R. An age-specific biokinetic model for iodine. J Radiol Prot 2017;37:864-82.Ferrari, et al. FLUKA: A Multi-Particle Transport Code CERN-2005–10, INFN/TC_05/11, SLAC-R-773. 2005.
Abstract - 42249: Internal dosimetry of workers handling 225Ac for targeted alpha therapy
I. S. Singh, M. Y. Nadar, Lokpati Mishra, Rajesh Sankhla, Deepak Akar, P. Mandal, P. D. Sawant
Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
The β/γ emitting radiopharmaceuticals such as 99Tc, 131I, 177Lu etc. are widely used for diagnosis and treatment of various kinds of cancer. In recent times several α emitting radionuclides viz. 211At, 225Ac, 213Bi, 227Th, and 223Ra have shown its importance in targeted therapy. Among these 225Ac/213Bi have emerged as most widely used α emitting radiopharmaceuticals for the treatment of malignancies.[1] The advantage of these radiopharmaceuticals is high linear energy transfer (LET) and ionization power of the emitted alpha particles as compared to the low LET β/γ emissions. The alpha-particle deposits higher energy in the malignant tissues and imparts less damage to the surrounding healthy tissues due to its low penetration power. Recently in India, Radiation Medicine Centre (RMC) Parel, has started using 225Ac for this purpose. 225Ac can be produced either by 226Ra (p, 2n) 225Ac reaction in a cyclotron or by chemical separation from 233U series. The alpha emitting 233U (T1/2 = 160,000 y) decays to 229Th (T1/2 = 7340 y), and chemically separated build-up activity of 229Th is used for the production of 225Ac/213Bi source. For radiological protection of the workers handling 225Ac, designing of individual monitoring program and interpretation of measurement results in terms of intake and committed effective dose (CED) is essential. Due to short half-life of 225Ac (10 h), proper planning to select the most appropriate bioassay methodology for the internal dose evaluation is required. In this paper, various internal contamination monitoring methodologies for workers handling 225Ac/213Bi sources are discussed. In-vivo or in-vitro methods [2] are mainly used for the measurement of internally deposited radionuclides. In-vitro monitoring requires around 3-4 working days in chemical separation, electroplating followed by α spectroscopy using PIPS detector. 225Ac and all its daughters are short lived, and, considering long analysis time for in-vitro methods, direct monitoring is more feasible for these kinds of short-lived radionuclides provided they emit penetrating x/γ photons. The important γ photons of 225Ac series are 217.6 keV (12.5% yield) and 440.4 keV (28% yield) from 221Fr and 213Bi, respectively and they are in secular equilibrium with 225Ac. In this work we have used 440.4 keV γ photon for direct measurements due to its higher energy and yield. The Quickscan, HPGe detector-based shadow shield bed whole body monitor (SSBWBM), and HPGe array inside steel room were evaluated for the in-vivo measurement of 225Ac. The efficiency of Quickscan and SSBWBM at 440.4 keV γ photon is evaluated from the empirical equations derived by distributing known amount of radio activities inside BOMAB phantom and performing efficiency calibration. The minimum detectable activity (MDA) for 225Ac is estimated as 430 and 320 Bq with Quicksan and SSBWBM for 5 and 30 min. monitoring time respectively. The efficiency of the HPGe array is evaluated from the empirical efficiency equation derived from the measurements of known amount of 152Eu uniformly distributed in the lungs of LLNL phantom. The MDA of HPGe array inside steel room for 225Ac deposited in lungs of worker is estimated as 12 Bq for 50 min. monitoring time. The minimum detectable intake and dose for all the three systems were estimated from the MDA values. The estimated CED of Quickscan and SSBWBM for MDA equivalent activity comes out above recording level (1 mSv). The MDA equivalent intake and CED of various absorption types of 225Ac are evaluated for HPGe array and given in [Table 1]. This shows, if measurements were carried out within 10 days after handling the absorption types M and S 225Ac, < 1 mSv CED can be estimated with HPGe array inside steel room. The results of this study show that lung monitoring in steel room is the best suited among the available options for internal contamination monitoring of 225Ac in radiation worker. This methodology is regularly used for internal dosimetry of workers handling 225Ac for targeted alpha therapy.{Table 19}
Keywords: Ac-225, internal dosimetry, in-vivo measurement, radiopharmaceutical, α-emitters
References
Zimmer MA. Vol. 11. Chicago, IL: Northwestern University Medical Center; 2004. p. 7-27.Pendharkar KA, et al. BARC Newsl 2008;296.
Abstract - 42251: Sequential separation of Plutonium and Americium in urine using DGA resin
Nanda Raveendran, Rupali Dubla, Ranjeet Kumar, Suja A. Kumar1, J. R. Yadav, Pramilla Sawant1
Health Physics Laboratory, IDS, RSSD, BARC, Tarapur, 1Internal Dosimetry Section, RSSD, BARC, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: The assessment of internal contamination of Plutonium (Pu) and Americium (Am) in occupational workers of nuclear facilities is carried out by bioassay analysis. In any incidental or accidental exposure condition, methods of detection have to be fast and sensitive enough to take remedial actions. There are few published methods in which DGA resin (N,N,N',N'-tetra 2-ethylhexyl diglycolamide) was used in tandem with TEVA/TRU resin for the sequential separation of actinides in urine samples.[1] In a work of Zagyvai et al., 2017,[2] single DGA resin column was used for actinides separation from an aliquot of urine sample. The present work is about sequential separation of actinides (Pu and Am) using DGA resin in one litre of urine sample.
Materials and Methods: Ten urine samples (1000ml each) of non-radiation workers were collected and used for standardization of the procedure. Samples were spiked with 242Pu and 243Am radionuclide sourced from National Physical Laboratory, UK (Reference No: E05080352, 2014110432-2 respectively) in the range of 1.6 - 5 mBq. These samples were wet oxidized with Conc. HNO3 + H2O2, followed by co-precipitation of calcium phosphate. The residue was dissolved in 10 ml of 3 M HNO3 containing 1 M Al(NO3)3 and 400 mg of NaNO2. This solution was loaded on DGA resin (0.6-0.8gm, 50-100 μ size) which was pre-conditioned with 30 mL of 3M HNO3.The column was washed with 20 ml of 3M HNO3 followed by 5 ml 2M HNO3. Initially americium was eluted from the column with 8 ml 0.1M HCl followed by plutonium elution with 15 mL of 0.1M oxalic acid in 1M HCl. The eluted Am and Pu fractions were electrodeposited and counted in alpha spectrometry for 86400 secs.
Extraction equilibria for Am+3, Pu+4 in DGA extractant can be given as follows:
[INLINE:2]
Results and Discussion: The analytical results of recovery are presented in [Table 1]. The radiochemical recovery of 243Am and 242Pu was obtained in the range 60-92%, 75-91% with a mean and standard deviation of 76.0% ± 11.3% and 85.0% ± 4.5% respectively. Minimum detectable activity of the method for Pu and Am estimation is 0.5 mBq for counting time of 86,400 secs. Some people[1] have used TEVA-TRU-DGA resin column for separation of Pu and Am in sample solution of 3M HNO3-1M Al(NO3)3 containing 1.5M sulfamic acid, 1.5M ascorbic acid and 3.5M sodium nitrite for Pu valency adjustment to IV. In the present method, single DGA resin column is used for separation of Pu and Am in urine sample. Sample load solution of the present work was prepared in 3M HNO3-1M Al(NO3)3 containing NaNO2 (400mg) for Pu valency adjustment. Addition of sulfamic and ascorbic acid were avoided without compromise in recovery. Study about sample elution from the column had shown that 8 ml 0.1M HCl and 15 ml of 0.1M oxalic acid in 1M HCl were adequate for complete elution of Am and Pu respectively.{Table 20}
Conclusion: Sequential separation of Pu and Am in urine samples using DGA resin and alpha spectrometry is presented. Use of single DGA column for actinides separation is convenient in handling and economical when compared with anion exchange and other published extraction chromatography work. The standardized method has resulted in separation of Pu and Am with an acceptable chemical recovery and it can be used as an alternative method for handling incidental/accidental exposure cases of radiation workers.
Keywords: Americium, DGA, plutonium, urine
References
Maxwell SL, Jones VD. Talanta 2009;80:143-50.Zagyvai M, Vajda N, Groska J, Molnar ZS, Bokori E, Szeredy P. J Radioanal Nucl Chem 2017;314:49-58.
Abstract - 42252: Screening of actinides in urine using Liquid Scintillation Counting technique
S. P. Prabhu1,2, S. P. Bhade1, M. S. Kulkarni2,3
1Radiation Safety Systems Division, Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, 3Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Timely assessment of internal contamination is important during radiation emergency for identifying contaminated individuals for early medical intervention. In the present study, a rapid technique was developed for estimation of actinides in urine samples using Liquid Scintillation Counting (LSC). The study comprises, coprecipitation of actinides as Ca3(PO4)2, dissolution of precipitate in dilute HCl medium followed by estimation using Pulse Shape Analysis (PSA) based LSC technique. Initially, study was carried out to select suitable acid medium for dissolution of Ca3(PO4)2 precipitate. For this study, to 500mL distilled water samples (6 nos.), 200 mg Ca-carrier, 5 mL H3PO4 was added. These samples were wet digested using conc. HNO3 and H2O2, and preconcentration was done by Ca3(PO4)2 precipitation. The precipitate was dissolved in 5 mL conc. HNO3 and wet ashed. Later HNO3 and HCl media with different molarities (1 M to 3M) were examined for complete dissolution of Ca3(PO4)2 ppt. 2M HCl acid medium was selected for dissolution of precipitate as clear and homogeneous solution was observed when 5mL of the sample was added to 10 mL of Optiphase Hisafe-3 scintillator in high density polyethylene (HDPE) vial. For the rest of the acid solutions, milky colour/ or phase separation was observed. LSC measurements were carried out using ultra low level Quantulus 1220. 500 mL urine samples (3 nos.) were processed to obtain Ca3(PO4)2 precipitate which was dissolved in 10 mL of 2M HCl. 5 mL of the sample aliquots in each of the three sets, were spiked with 233U α-standard (15.3 Bq), 90Sr/90Y β-standard (27.7 Bq) respectively while 3rd vial with no activity added was used as blank. These α and β standards along with the blank were used for calibration of LSC. In these standards, various quench levels (530-700) were simulated by adding Nitromethane in incremental amount. PSA settings were optimized for various quench levels and subsequent counting and spillover efficiencies were obtained.[1] Based on the blank sample analysis, the MDA achieved for alpha activity is 10.8 mBq/500 mL for 1h counting time which is less than the derived alpha activity in urine corresponding to dose of 0.1 Sv [Table 1]. 0.1 Sv is the annual effective dose limit for emergency situations.[2] The proposed method was validated by analysing spiked urine samples (10 nos.). Known amount of α and β activities were added to urine samples in various α:β ratios 1 to 1:50, 2:1 and 10:1. This was done to check the accuracy of the standardized method for estimation of α-activity in presence of β-activity. The observed α-activity linearly agreed well with the spiked values [Figure 1]. The relative bias of all spiked urine samples was within ± 15%. Discriminated α and β LSC spectra of spiked urine sample is shown in [Figure 2]. The procedure standardized is rapid and will be useful for estimation of actinides in urine samples during radiation emergency. Total time required for the standardized method including counting is 1 working day.{Table 21}{Figure 39}{Figure 40}
Keywords: Actinides, calcium phosphate, liquid scintillation, urine
References
Bhade SP, Reddy PJ, Narayanan A, Narayan KK, Babu DA, Sharma DN. J Radioanal Nucl Chem 2010;284:367-75.ICRP. ICRP Publication 103. Annals ICRP 37; 2007. p. 1-332.OIR-4. ICRP Publication 141. Annals ICRP 48; 2019. p. 9-501.
Abstract - 42259: Calibration coefficients for monitoring of uranium and thorium in lungs
C. S. Charubala1, S. L. Asanali1, V. Santhanakrishnan1, G. Ganesh1, M. S. Kulkarni1,2
1Health Physics Division, Bhabha Atomic Research Centre,
2Homi Bhabha National Institute, Mumbai,
Maharashtra, India
E-mail: [email protected]
HPGe based detection systems are used for the in-vivo monitoring of radiation workers to detect the internal contamination due to low energy photon emitters such as actinides. An array of three HPGe detectors commissioned in a steel room and calibrated using LLNL physical phantom has been used for the routine internal monitoring purposes in our laboratory. For the monitoring of uranium in lungs, the detection system is calibrated using LLNL phantom with indigenously made natural uranium pellets distributed in the lung mould in a way to simulate the uniform distribution.[1] In the current study, this uranium tagged phantom was positioned under the detection system according to the measurement geometry and the spectra were acquired for various chest wall thicknesses. Calibration coefficients (η) were determined considering source activity, counting time and gross counts subtracted from background for each regions of interest. In FLUKA simulations,[2],[3] BEAM card was used to define interested energies, 63keV and 93keV (234Th from 238U), 74keV and 238 keV (212Pb from 232Th) and 186keV (235U). A source routine was used for sampling the lung voxels for uniform distribution of sources. DETECT card was used to score counts/photons which was further converted to cps/Bq after considering the yield. The simulations were performed for not less than 5x106 histories to reduce statistical error less than 3%. The Muscle Equivalent Chest Wall Thickness (MEQ-CWT) values corresponding to the ICRP-AM thorax voxel phantom for various energies were calculated by voxel counting method. For this, adipose muscle fraction of 48% and linear attenuation coefficients (as derived from mass attenuation coefficients from NIST data base after interpolating to desired energies by plotting and density) were also used. η values from experiment and simulations were compared for the MEQ-CWTs of ICRP-AM for various energies. The uniform distribution of source (particle scoring) in lungs and spectra for various energies from FLUKA simulations are illustrated in [Figure 1]. The calibration coefficients corresponding to MEQ-CWTs for ICRP-AM obtained from experiment and simulations are shown in [Table 1]. Since thorium lung set was unavailable, η values for corresponding energies are derived from only simulations. The uncertainties in experimental values in [Table 1] refer to the combined uncertainty due to counting statistics and source activity. It can be seen from [Table 1] that η for 63keV differs by <1%, 93keV differs by 6% and 186keV differs by 25%. Even though the similarities of this uranium tagged LLNL phantom with uniformly distributed lungs source for 63 keV and 93 keV was previously established in the original publication of this phantom, the similarity for 186 keV was not proven by any previous research. However, the current study finds significant deviation for using the uranium tagged LLNL phantom in calibration of the HPGe array detection system for 186 keV gamma rays.{Figure 41}{Table 22}
Keywords: Calibration, FLUKA, in vivo monitoring, whole body counting
References
Manohari M, Deepu R, Mathiyarasu R, Rajagopal V, Jose M, Venkatraman B. Calibration of Phoswich detector for the measurement of natural uranium in lungs. Radiat Prot Environ 2016;39. [doi:org/10.4103/0972-0464.190391].Ahdida C, Bozzato D, Calzolari D, Cerutti F, Widorski M. New capabilities of the FLUKA multi-purpose code. Front Phys 2022;9.Vlachoudis V. Flair: A Powerful but user Friendly Graphical Interface for FLUKA; 2009. p. 2.
Abstract - 42264: Optimization of strontium carrier addition for the rapid estimation technique of 90Sr in urine
C. P. Jayasree1, Thamizharasi Ganesh2, V. Santhanakrishnan1, G. Ganesh1, M. S. Kulkarni1,3
1Health Physics Division, BARC, 2INRPK, Kalpakkam,
Tamil Nadu, 3HBNI, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: In the event of radiological emergencies, rapid radio bioassay program provides information about internal contamination for further medical management of radiation workers as well as the public. 90Sr is an important fission product which emits beta particle with a half life of 28 years. Extraction chromatographic (EC) technique can be used for rapid separation of 90Sr in urine samples.[1] In this study, standardization of procedure for rapid separation and estimation of 90Sr in urine samples was carried out by minimizing sample volume and optimizing the carrier quantity of stable strontium.
Materials and Methods: 20 mL aliquots of urine sample collected from non radiation worker was used in this study.[2] Sr carrier was added to these aliquots prior to wet ashing of the sample with conc.HNO3 and H2O2 and the sample was evaporated to dryness. The residue was dissolved in 5mL of 8 M HNO3 and was loaded in EC column containing 3mL bed volume of preconditioned Sr specific crown ether resin. Sr was stripped using 0.05 M HNO3 and precipitated as SrCO3 using saturated Na2CO3 at pH9 and digested for 30 minutes in a water bath. The precipitate was deposited on pre-weighed glass fiber filter paper (1 μm pore size, 25 mm dia), dried and recovery was calculated gravimetrically. Sr carrier concentration was optimized by varying the concentration of Sr carrier (2.7 to 34 mg) and estimating the Sr carrier recovery. Radiochemical recovery study was done by analysis of 20mL aliquot samples (20nos.) spiked with known activity of 90Sr (0.5-15Bq) and the optimized concentration of Sr carrier. 90Sr activity was estimated using a low background beta counting system.
Results and Discussion: 54 urine samples were analyzed by adding different concentrations of Sr carrier (2.7-34 mg) for optimizing the carrier weight. It was observed that addition of 5-7 mg of Sr carrier was sufficient for a 3mL bed volume of resin to achieve chemical recovery of 75-100% as shown in [Figure 1] and with radiochemical recovery above 85%. Recovery of Sr carrier at higher concentrations is limited by the capacity factor of the resin which is usually 10 to 20% of the maximum capacity of 10 mg per mL of bed volume. Minimum detectable activity (MDA) of this method was found to be 63 mBq for 20 mL sample (3600 seconds counting) which is lower than the action level of 64 Bq[3] for emergency screening of 90Sr in urine samples. Relative bias and precision for this method calculated as per ISO 28218 criteria are well within the prescribed limits. Summary of the study is given in [Table 1].{Figure 42}{Table 23}
Conclusion: Rapid emergency screening for 90Sr intake with EC technique using Sr resin is feasible. Good radiochemical recovery with minimum sample volume and optimum Sr carrier addition has been determined.
Keywords: 90Sr, rapid estimation, urine
References
Horwitz PE, Chiarizia R, Dietz ML A novel Strontium selective extraction chromatographic resin. Solv Extrn Ion Exch 1992;10.Maxwell SL, Culligan BK. New column separation method for emergency urine samples. J Radioanal Nucl Chem 2009;279:105-11.Wankhede Sonal M, Sawant Pramilla D, Rao DD, Pradeep Kumar. Radiat Prot Environ 2014;37.
Abstract - 42266: Uncertainty in the estimation of intake of actinides due to unknown time of intake
M. Y. Nadar, S. Halder, L. Mishra, I. S. Singh, P. D. Sawant
Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
In routine monitoring for internal contamination, when time of intake is unknown, it is assumed that intake has occurred at the midpoint of monitoring interval.[1] An uncertainty[2] in intake estimation is introduced due to this assumption as time of intake can vary from 1 to 365 day for annual routine monitoring frequency. Radiation worker is monitored at the end of monitoring interval, within a span of ± 1 month. In this study, uncertainties introduced in the estimation of intake of type M and S compounds of actinides due to unknown time of intake are evaluated. Lung retention data of revised Pu biokinetic model given in OIR, part 4[3] is used to evaluate the uncertainty. In this study, we have used minimum detectable activity (MDA) of lung monitoring system using HPGe array of 5 Bq of 241Am. 241Am is used as a tracer to detect Pu in lungs when it is in Pu matrix. A python program is written to randomly choose a day between first and last day of monitoring interval. For the randomly chosen intake day, three probable monitoring days at the end of the interval in the span of ± 1 month were estimated. Retention fraction at these days is taken to estimate the corresponding intake. For an assumed day of intake, d, three monitoring data points, such as 335-d, 365-d and 395-d are obtained for annual monitoring interval (365 day). Similarly, for two-year monitoring interval, three data points, such as 700-d, 730-d and 760-d are considered and intakes are evaluated using respective retention fractions. For annual monitoring interval and assuming intake at midpoint, the expected intake to 5 Bq of 241Am lung activity is 157 Bq. For biennial monitoring interval, the corresponding intake is 188.7 Bq. From the evaluated intakes at randomly selected 365 days, the average intake is found to be 157.5 ± 23.5 Bq assuming Gaussian distribution. Assuming log-normal distribution, the median value obtained is 155.5 Bq with scattering factor (SF) of 1.18. The frequency distribution obtained for lung activity estimation is given in [Figure 1]. It is found that unknown time of intake does not introduce relative error more than 15 % in lung monitoring results, for monitoring interval of 1 year for type S compounds. Thus, it is most robust and accurate method for assessment of internal contamination with annual monitoring interval when time of intake is unknown. The uncertainty also studied for biennial monitoring interval for type S compounds. From the evaluated intakes at randomly selected 365 days from 1 to 700 days, the average intake is found to be 195.4 ± 57.2 Bq assuming Gaussian distribution. Assuming log-normal distribution, the median value obtained is 201.2 with SF of 1.30. The frequency distribution obtained for lung activity estimation is given in [Figure 2]. It can be seen from [Figure 2] that spread is increased for biennial monitoring interval compared to annual monitoring interval. The observed mean intake is 5 % higher than the expected intake with relative error of 30 %. The results of this study show that for insoluble type S compounds, if monitoring interval is varied from 1 to 2 years, SF due to unknown time of intake varies from 1.18 to 1.3. Similar study was carried out for type M compounds and SF is found to be 2.06 for annual monitoring interval.{Figure 43}{Figure 44}
Keywords: Pu/Am, scattering factor, uncertainty, unknown time of intake
References
ICRP. Individual monitoring for internal exposure of workers. ICRP Publication 78. Ann ICRP 1997;27.NCRP Report 164. Uncertainties in Internal dose Assessment; 2009.Occupational intakes of radionuclides: Part 4. ICRP Publication 141. Ann ICRP 2019;48.
Abstract - 42348: 226Ra body burden among smoker and non- smokers group of Uranium Mill workers of Jaduguda, India
M. K. Singh1,2, R. L. Patnaik1,2, V. N. Jha1,2, D. Rana1,2, S. K. Jha2,3, M. S. Kulkarni2,3
1Health Physics Unit, RPS(NF), HPD, BARC, Jaduguda, Jharkhand, 2Health Physics Division, BARC, 3HBNI, Mumbai, Maharashtra, India
E-mail:[email protected]
Introduction: In uranium milling facility workers are exposed to uranium and its daughter products during different operations such as crushing, grinding, screening, dissolution and finished product separation. These radionuclides with diverse radiological (half life, radiation type and energy) and physicochemical features form potential source of exposure of workers. Among the radionuclides of uranium series 226Ra assume significance due to long radiological half life, high alpha energy and associated bone toxicity (chemical similarity with Ca).The radionuclide find their way into the metabolic system mostly through the inhalation route. The gaseous decay product 222Rn is released through the lung during gaseous exchange and measurement can reflect the 226Ra content of the subject (occupational worker).[1] The physiological response of particulate inhalation is quite complex, particularly if the subject has other related habit attributes. Impact of cigarette smoke has been globally studied and compromised lung efficiency is a common feature of smoker. Present investigation has been carried out to compare the 226Ra body burden among two groups of occupational workers of uranium ore processing unit at Jaduguda classified into smoking and non-smoking (control) categories.
Materials and Methods: Radon-in-breath measurements are carried out for the estimation of 226Ra body burden.[1] The subject inhales radon free medical oxygen through a respirator, and the exhaled air is led to a low-level radon detection system chamber (LLRDS) after passing through an ice-cold moisture trap. The 222Rn present in the collected samples decays to 218Po, which is positively charged at the time of sampling. Following the sampling after specified delay the 222Rn concentration is evaluated via counting the alpha activity of the daughter 218Po and subsequently the 226Ra body burden 'q' is obtained using Eq. 1using the ( Srivsatava 1986).
[INLINE:3]
where, CRn = Radon concentration in breath air (Bqm-3), Vb = Breathing rate (1.08 x 10-4 m3s-1), f = fraction of 222Rn exhaled (0.7), l = decay constant of radon(s-1).
Results and Discussion: Different job categories and habitat of about 155 workers from uranium ore processing facility has been monitored for radon-in-breath and their 226Ra body burden was estimated using equation 1. Results are compiled for the period 2015 to 2020 and provided in [Table 1]. The 226Ra body burden of workers of all category ranged from 0.12- 3.99 kBq with an average of 2.18 ± 0.96 kBq. Among these results average 226Ra body burden of workers for chemical group and maintenance job category of processing plant was 2.27 ± 0.95 kBq and 2.01 ± 0.96 kBq respectively. The classification of these workers in smoking and non smoking category was observed to be an average 226Ra body burden of 2.63 ± 0.94 kBq and 2.04 ± 0.92 kBq respectively. The preliminary findings indicated elevated Radium body burden among smokers as compared to the non smokers. Although, in both the categories the 226Ra body burden is well below the ALI of 9.2 kBq. The variability in the 226Ra body burden among two groups can be associated with the physiological aspects, lung efficiency, breathing rate and additional input from particulate matter originating from tobacco smoke.{Table 24}
Conclusion: Elevated level of 226Ra body burden among smokers category of mill workers was observed in comparison to non smokers based on this preliminary study. The study may be continued further for uranium ore processing workers for future investigations on contributing factors like smoking habits etc.
Keywords: 226Ra, body burden, ore processing, smokers
References
Patnaik RL, Jha VN, Kumar R, Srivastava VS, Ravi PM, Tripathi RM. Distribution of 226Ra body burden of workers in an underground uranium mine in India. Radiat Environ Biophys 2014;53:739-44.Srivastava GK, Raghavayya M, Kotrappa P, Somasundaram S. Radium-226 body burden in U miners by measurement of RN in exhaled breath. Health Phys 1986;50:217-21.
Abstract - 42366: Development of an indigenous software 'IDES-V1.0' for radionuclide intake estimation
S. Halder, D. Akar, M. Y. Nadar, P. K Singh, A. D. Otari, J. R. Yadav, P. D. Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
In nuclear industry, occupational worker may get internally contaminated via inhalation, ingestion or wound. The committed effective dose (CED) is estimated using intake which in turn depends on various parameters such as date of intake, uncertainties in sample analysis, measurement techniques etc. Statistical methods such as, maximum likelihood, least square & Bayesian are used to analyze the measurement data & estimate most probable intake. Currently, IMBA software (Birchall et.al., 2007)[1] is used for bioassay data analysis which implements the ICRP biokinetic and dosimetric models. The model parameters can be modified by the users to obtain the best fit for the observed monitoring data. The IMBA software is compatible with Windows based operating system. In this study, development of in-house software, Internal Dose Estimation Software (IDES-V1.0) is discussed. The software estimates intake in single and/or multiple acute exposure scenarios. It analyzes data obtained using single bioassay measurement technique. In the present version, computations are performed using maximum likelihood method. For data having log normal distribution and constant scattering factor, intake can be estimated using following equation:[1]
[INLINE:4]
Where, Mi and mi are the measured data and excretion/retention fraction, respectively and each Mi/mi is intake estimated using ith measurement. The calculation methodology is independent of biokinetic model and provides user a flexibility to input mi values derived using suitable biokinetic model. In case of multiple intake scenarios, the data is corrected for contributions from previous intakes during analysis & all intakes are calculated one by one separately using the above equation. In the software, modules for data analysis have been implemented using python while graphical user interface (GUI) of the software is developed in visual studio [Figure 1]. Software can handle 119 types of data combinations such as- exposure information (F, M & S type of material), routes of intake (inhalation, ingestion, and injection), type of measurement data (Urine, Fecal, Lung, Whole body etc.), radionuclides etc. Software provides graphical representation of measurement data and its fitting pattern, which are used to make important decisions, such as probable number of intakes, date of intake & type of material (F, M & S). Statistical analysis of data provides better fitting results which have acceptable chi-square, auto-correlation coefficient and p-values. Results are stored in a file with applied input parameters and statistical analysis. The software was validated by solving a few of the IAEA intercomparison cases. Intake and CED values of two such cases estimated with IDES and corresponding IAEA values are shown in the [Table 1]. Chi-square, auto correlation and p-value for Case-1 are 14.4, 0.86 and 0.88 respectively, For Case-7 these values are 16.9, 0.50, 0.39 respectively. Based on these values, estimated results are accepted. From [Table 1], it can be seen that the differences in results of IDES and IAEA values are < 0.5%. IDES software has been developed and validated by solving bench marks cases used in the IAEA international intercomparison exercise. IDES V-1.0 has passed alpha test and beta testing is in progress. In its newer version statistical tools such as least square and Bayesian analysis will also be incorporated along with flexibility to choose in old and latest biokinetic/dosimetric parameters.{Figure 45}{Table 25}
Keywords: Chi-square and auto correlation, IDES, multiple intakes
References
Birchall A, Puncher M, Marsh JW, Davis K, Bailey MR, Jarvis NS, et al. IMBA professional plus: A flexible approach to internal dosimetry. Radiat Prot Dosimetry 2007;125:194-7.IAEA. Intercomparison Exercise on Internal Dose Assessment. IAEA; 2007.IAEA. Intercomparison and Biokinetic Model Validation of Radionuclide Intake Assessment. IAEA; 1998.
Abstract - 42378: Adequacy of direct lung burden measurement for different type of Pu intake
J. R. Yadav, P. K. Singh, Smita Thakur, Rahul Roy, Soumitra Panda, Pramilla Sawant
Internal Dosimetry Section, Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Direct lung burden measurement of 239Pu in occupational workers, presents difficulties due to attenuation of low energy & low yield photons (Eγ =17keV (4.6%)) by overlying tissues of chest. Further, ultra sound and NMR imaging can be used for the measurement of chest wall thickness for incidental intake cases. 239Pu lung burden is usually extrapolated from the measurement of 60 keV photon emitted by 241Am formed by the decay of 241Pu present in the inhaled material. Currently, in H.P. Lab, two HPGe detectors of crystal dimension 91.3 mm (dia) x 31.3 mm (thick) having thin carbon epoxy window 1 mm (thick), installed inside steel room having graded shielding, are used for Pu lung burden measurement.[1] The system is calibrated experimentally using realistic tissue equivalent thorax phantoms such as Lawrence Livermore National Laboratory (LLNL) for 241Am. Minimum detectable activity (MDA) of system for 241Am lung measurement is 7 Bq. Annual limit of intake (ALI) for 239Pu based on ICRP-66 recommendations are 625 and 2400 Bq for type M and S respectively. The present study highlights the data which can be used as deciding factor for direct lung burden measurement of Pu.
Materials and Methods: Composition of plutonium isotopes (239Pu; 241Pu) in PHWR nuclear power plant varies depending upon fuel burn up. For this study, published data for plutonium burn up[2] along with Bateman equation[3] was used for the assessment of 241Am in growth from 241Pu present in the spent fuel.
[INLINE:5]
Where λA is decay constant for 241Pu and λB is decay constant for 241Am. 239Pu/241Am ratios (hereafter referred as Pu/Am) in spent fuel are assessed over a period of 5 years knowing composition of 239Pu from published data and 241Am in growth calculated from Eq 1.
Results and Discussion: [Table 1], presents measurable intake in terms of ALI values of Pu based on 241Am measurement of the system for acute (1 day) and routine monitoring scenarios (180 days) for type M and S respectively. For Pu/Am ratio of 3.6 (cooling period 5 Y), 0.7 & 0.16 ALI values are achievable for type M & S respectively. When the Pu/Am ratio is 10 (cooling period 1.8 Y), which is generally assumed in reprocessing plant, 1.90 & 0.45 ALI intake of type M & S of Pu are achievable respectively. During routine measurement present system can measure 3.4 ALI of M and 0.33 ALI of type S Pu intake for 5 yrs cooled spent fuel. But for short cooled fuel (1.8 y) ∼10 ALI of type M and 0.9 ALI of type S intake can be assessed.{Table 26}
Conclusion: Assessment of Pu lung burden for type M acute intake and < ALI is possible through measurement of its tracer radionuclide (241Am) beyond 3 y cooled fuel. However, for type S intake this measurement is possible even for 1y & beyond cooled fuel. For routine monitoring of type S Pu-intake, assessment of intake for <1 ALI can be done when fuel is cooled beyond 1.8 y. However, same is not possible for type M intake. Therefore for type M compound, in vitro bioassay i.e. urine and faecal measurement for intake < ALI can easily be carried out. Faecal bioassay has duel advantage: it is useful in assessing type M and S low level intakes (≤ 0.1 ALI) & it can also provide information in deriving effective activity median aerodynamic diameter (AMAD) when 241Am is also measurable by lung counting.
Keywords: 239Pu/241Am ratio, bateman equation, lung measurement
References
Pradeep, et al. Development of HPGe Detector Based In-vivo Monitoring System for Workers Handling Actinides, NUCAR-2021, February 22-26; 2022. p. 350.Alamelu, et al. BARC New Letter Issue No. 313, March-April 2010.Jerzy, et al. General solution of Bateman equations for nuclear transmutations. Ann Nucl Energy 2006;33.
Abstract - 42548: Studies on the effective half-life of elemental tritium uptake through inhalation pathway and its relation with ambient climatic conditions
V. Narasimha Nath, C. N. Sunil, V. Ramakrishna, G. Ganesh, M. S. Kulkarni
Health Physics Division, BARC, Mumbai, Maharashtra, India
E-mail: [email protected], [email protected]
On an average, about 20 % of the radiation dose received by a radiation worker in a Pressurised Heavy Water Reactor (PHWR) can be assumed to be contributed by tritium (3H), which is an activation product of heavy water (D2O) used as the coolant and moderator of PHWR. In PHWR, tritium is present in aqueous & vapor form (DTO, HTO and T2O) as well as in elemental form (HT, DT, T2) due to radiolysis. Even though the internal dose monitoring program is ardently followed in these facilities, it is for the first time in India, a study was undertaken for estimating the elimination rate of elemental tritium uptake from human body. Tritium concentrations in urine were estimated using Hidex 300SL Liquid Scintillation Analyser equipped with Triple to Double Coincidence Ratio (TDCR) method having MDA of nearly 1 Bq/ml. For pure beta emitters like 3H, TDCR value numerically equals with counting efficiency. Under this study, urine samples from 34 personnel with significant tritium uptake were analysed. Out of 34 cases, based on the nature of job 30 had exposure only in elemental form and three had vapour uptake which are given in [Figure 1] and [Table 1] respectively. The activity trend was followed in all cases till it reached below recording level and established the durations for successive effective half-lives (TE). The mean TE is about 7 days in vapor form which is almost at par with the uptake in elemental form. None of the personnel were given any diuretics or advised for during the entire course of observation. Different studies report the effective half-life of HTO varies from 4-18 days.[1],[2] The ICRP's HTO model[3] adopted a biological half-life of 10 days for adults.[2] Butler and Leroy found that effective half-life varies with water intake,[4] ambient temperature (decreasing with increasing ambient temperature) and age (decreasing with increasing age in adults).{Figure 46}{Figure 47}{Table 27}
Conclusion: All the cases in this study showed average biological half-life in three different climatic conditions varying from 2.8 to 9.64 days. Effect of climatic conditions on biological elimination of HTO is clearly seen. The study showed that the elimination rate is almost similar for both elemental and vapor forms of tritium uptake.
Keywords: Effective half-life, inhalation, tritium
References
Hill RL, Johnson JR. Health Phys 1993;65:28-647.Rudran K. Radiat Prot Dosimetry 1998;25:5-13.Annals of ICRP (Publications no. 67, 71, 134).Butler HC, Leroy JH. Health Phys 1965;11:283-5.
Abstract - 42608: Study on detection efficiency response of standing whole-body counter based on multi-size voxel phantoms
Xiaodun Li, Yunshi Xiao, Qinjian Cao, Yuan Zhao, Liye LIU
Department of Health Physics, China Institute for Radiation Protection, Taiyuan, China
E-mail: [email protected]
The standing whole-body counters are widely used for in-vivo measurements, such as ORTEC-StandFAST II, CANBERRA-FASTSCAN, CIRP-StandWBC. They are mainly designed to quickly and accurately measure internal contamination of radionuclides with medium and high photon energy. The physical phantoms such as BOMAB (Bottle Manikin Absorption Phantom, ANSI/HPS N13.35-2009), IGOR(also called unified phantom, developed at the Research Institute for Industrial and Marine Medicine and the Research and Technical Centre Protection, St Petersburg), sBCAM (solid BOMAB phantom for Chinese Reference Adult Male, CIRP) have been used for whole-body counting calibrations. However, these simplified phantoms cannot fully represent all the characteristics of the human body, furthermore, the phantom parameters are based on the reference human body, which cannot represent workers with different body shapes and sizes. To date, there is a lack of understanding about the detection efficiency response as a function of phantom sizes and photon energy. This paper presents a study on the detection efficiency of standing whole-body counter obtained from Monte Carlo simulations of multi-size voxel phantoms. The Monte Carlo method was used to simulate and analyze the detection efficiency response of standing whole-body counter to multi-size voxel phantoms. The whole-body counter is ORTEC-StandFAST II. The multi-size voxel phantoms are Chinese adult male individualized voxel phantoms, with a height range of 155cm to 185cm and a weight range of 42 kg to 103 kg; The photon energy range is 80 keV to 1836 keV; And radioactive substances are homogeneous distribution through the whole body. In this paper, the body build index (BBI, BBI = (W/H)1/2 is introduced to analyze the influence of human body size on the detection efficiency response of whole-body counter. Compared with the detection efficiency of the calibration phantom, the relative deviation range of the detection efficiency response of the whole-body counter to other sized voxel phantoms is -20.15% to 33.03%. A function was also found that related detection efficiency to BBI and photon energy. Based on this function, the deviation of detection efficiency between the calculated value based on this function and the simulated value is within -4.41% to 8.15%. This paper presents a study on the detection efficiency of standing whole-body counter obtained from Monte Carlo simulations of multi-size voxel phantoms. The results show that the detection efficiency of whole-body counter largely depends on the phantom size, and the function can be used to modify the efficiency calibration results for a given individual, which can effectively improve the measurement accuracy of the whole-body counter.{Figure 48}{Figure 49}
Keywords: BBI, efficiency calibration, multi-size, voxel phantom, whole-body counting
Abstract - 42609: Preliminary study on the effect of DTPA on plutonium transfer in liver cells
Cui Shuangshuang, Dong Juancong, Cheng Jiao, Liu Hongyan, Li Youchen
China Institute for Radiation Protection, Taiyuan, Shanxi, China
E-mail: [email protected]
Plutonium is one of the main occupational hazards for reprocessing professionals, and it is also the focus of radiation protection. In order to assess the degree of harm to occupational personnel, it is necessary to estimate the internal exposure dose. After plutonium exposure, DTPA is often used to promote the excretion of plutonium, but the use of DTPA will affect the plutonium biokinetics and thus affect the internal dose estimation results. At present, the research on liver biokinetics of plutonium mainly focuses on data analysis of professionals and extrapolation of animal experiments to humans.[1],[2],[3] However, the differences between individuals and species lead to great uncertainty in the data. At present, studies have been conducted on plutonium transfer using an in vitro cell model, but mainly for respiratory tract, and the plutonium exposure dose used is high (10kBq/106 cells).[4] In this study, we conducted a study on the transfer of plutonium in liver cells by DTPA through plutonium exposure in vitro, observed the impact of DTPA on plutonium excretion from liver cells, and provided data support for the estimation of internal exposure dose and a new idea for plutonium cell biokinetics research. Normal human liver cell line (L-02) was cultured in plutonium(242Pu) containing medium (0.097Bq/106 cells) were cultured for 24h, and the plutonium content in the cells was measured by ICP-MS. The detection limit of 242Pu by ICP-MS can reach 0.1pg/mL. After optimization, the current measurement process includes adding plutonium to the liver cells after 4 days of culture, collecting and counting liver cells, digesting cells with concentrated nitric acid, and measuring plutonium content by ICP-MS. After optimizing the experimental process, the measurement error of ICP-MS was reduced from 16% to less than 5%, meeting the experimental requirements. The poisoned cells were cultured under +/-DTPA(0.5mM) for 2h, 4h, 8h, 24h and 48h respectively, and the plutonium content in the cells was measured. The excretion of plutonium by liver cells exposed to toxin at different time points was observed. It was found that the plutonium content in hepatocytes without DTPA decreased from initial 89.750 ± 2.662pg/106 cells to 31.530 ± 2.189pg/106 cells at 48h, a decrease of 64.869%, while that in DTPA group decreased from 89.750 ± 2.662pg/106 cells to 21.917 ± 1.862pg/106 cells at 48h, a decrease of 75.580%. In this study, it is confirmed that DTPA reduces the plutonium content in cells by using the detection method of trace plutonium in liver cells, and the ratio of plutonium content between the DTPA free group and the DTPA group was 1.256 ± 0.177.
Keywords: ICP-MS, liver cells, plutonium
References
Leggett RW, Eckerman KF, Khokhryakov VF, Suslova KG, Krahenbuhl MP, Miller SC. Mayak worker study: An improved biokinetic model for reconstructing doses from internally deposited plutonium. Radiat Res 2005;164:111-22.Durbin PW. Plutonium in man: A new look at the old data. Radiobiol Plutonium 1972.Warner AJ, Talbot RJ, Newton D. Deposition of plutonium in human testes. Radiat Prot Dosimetry 1994;55: 61-3.Van der Meeren A, Moureau A, Laurent D, Laroche P, Angulo JF. In vitro assessment of plutonium uptake and release using the human macrophage-like THP-1 cells. Toxicol In Vitro 2016;37:25-33.
Abstract - 42611: Development of a new skull phantom for calibrating in vivo monitoring systems for Pb-210 internal contamination
Liu YiCong, Xiong WanChun, Xiao YunShi, Li XiaoDun
China Institute for Radiation Protection, Taiyuan, China
E-mail: [email protected]
The internal exposure dose estimation formula is E = I·e(g), where I is the intake, and e(g) is the dose coefficient. Therefore, the intake I directly determines the committed effective dose, and the main purpose of internal exposure monitoring is to obtain the value of the intake I by measuring the retained activity In order to improve the accuracy of the intake I, a more effective measure is to calibrate the monitor to improve the measurement accuracy. In this paper, a new anthropometric skull phantom has developed and applied for calibration of skull counter. The phantom's head circumference of is 56.18 cm, the total head height is 23.31 cm, the maximum head breadth is 15.76 cm, and the maximum head length is 19.23 cm which conform to the reference Chinese male.[1] The phantom consists of bone substitute and soft tissue substitute, the soft tissue substitute is polyurethane, and the hard bone substitute is a mixture of epoxy resin and calcium carbonate. At 46.5keV, the relative deviations between the mass attenuation coefficients of the two tissue equivalent materials and the reference values given by the ICRP[2] are 0.86% and -2.98%, respectively. The relative deviation of the mass attenuation coefficient of brain tissue and soft tissue at 46.5 keV is 5.03%, so soft tissue substitute can be used instead of brain tissue. The phantom was made by pouring. The mold is made by 3D printing. In order to restore the bone-seeking nuclide distribution, the authors decided to simulate the surface source by evenly distributing point sources on the skull surface. The solution of the radioactive source was dropped on the filter paper, and after drying the filter paper was cut and attached to the corresponding position on the skull, and then the skull was suspended in the head mold. After the positioning was completed, soft substitute was poured and the radioactive sources sealed. The total Pb-210 content in the point sources is 4203 Bq. The skull counter is a HPGe with the model of GEM40P4-76-S. The entrance window made of carbon. The crystal's radius is 40 mm and length is 30 mm. The measurement was carried out in a low-background laboratory, which is is 5 m deep, surrounded by 0.5m concrete, embedded with 0.4 m steel plates inside. The core part is composed of 20 cm thick iron plates. The internal space is 2.3 m long, 1.8 m wide and 2.2 m high. The dose rate in the core part is 26 nGy/h. In order to ensure the consistency of multiple measurements, a limit device was designed for the phantom. The detector plane was close to the scalp during measurement. The average detection efficiency after three replicate measurements, was 3.07 × 10-4cps/Bq. When the measurement time is 15 min, the MDA is 378 Bq. In this paper, an anthropometric skull phantom that conforms to the reference Chinese male is designed and manufactured. Using this model, the efficiency of the skull counter was evaluated, for accurate acquisition of Pb-210 retained in the radiation worker.{Figure 50}{Figure 51}
Keywords: Anthropometric phantom, calibrating, in vivo measurement
References
Bai X, et al. GBZ/T 200.1-2007 Reference Persons for Radiation Protection – Part 1: Physical Parameters.ICRP. ICRP Publication 23 Report of the Task Group on Reference Man. Oxford: Pergamon Press; 1975.
Abstract - 42626: Construction of dose-response calibration curves for micronuclei Induced by X-ray
Qianqian Meng, Zhongxin Zhang, Yue Ren, Jiao Cheng, Xiaozhen Li, Jingjie Wang, Ruifeng Zhang, Xuhong Dang
China Institute for Radiation Protection, Taiyuan, Shanxi, China
E-mail: [email protected]
In order to estimate dose by biodosimetry, any biological dosimetry service must have its own dose response calibration curve. Radiation-induced micronuclei (MN) can be used as a biological dosimeter for radiation protection because of the radiation quality and dose-dependence. To correlate the induced MN with the absorbed dose of the individuals, a reliable dose–response calibration curve should be established. This study aimed to investigate the MN frequency of human peripheral blood lymphocytes (PBLs) after exposure to different doses of X-ray, and to establish a dose-response relationship that would be highly appropriate for laboratory conditions of this study. A cytoge-netic study was conducted by use of the cytokinesis-blocked micronucleus (CBMN) assay. PBLs from nonsmoking male healthy donors who aged 25-33 (mean age, 27) years old were irradiated with various doses ranging from 0 to 5 Gy with 6MeV X-rays at a dose rate of 1Gy/min and scoring by the MN in binucleated (BN) cells. CABAS and Dose Estimate software were used to fit the MN and dose into a linear quadratic model, and the results were compared. The number of binucleated cells scored for the control group was 13000, 5000 cells for 3 Gy, 4000 cells for 4 Gy and 5 Gy, and 6000 cells for other dose points. Data from all donors were pooled together and frequency and distribution of MN in BN cells were analyzed. The results showed that baseline MN was 18.85 ± 1.20/1000 BN cells, and the MN frequency increased dose-dependently. At the same time, MN data revealed an over-dispersion trend at all doses, deviating from the Poisson distribution. With taking into account the pooled data of the three donors, the resultant curve calculated by the CABAS software was fitted to Y = (0.01772±0.00099) + (0.05999 ± 0.00331) D + (0.02989 ± 0.00097) D2. While by the Dose Estimate software, the fitted curve was Y = (0.0177 ± 0.0042) + (0.0603 ± 0.0133) D + (0.0298 ± 0.0037) D2. Both the two curves matched well. The linear and quadratic coefficients obtained by the two software were basically the same, and were comparable with published curves of similar radiation quality and dose rates by other studies, despite a certain degree of variation. In conclusion, The dose-response calibration curves for X-ray-inducded MN in human PBLs have been firstly established in our laboratory, which can be used as an alternative method for in vitro dose reconstruction and provides a reliable tool for biological dosimetry in accidental or occupational radiation exposures.{Figure 52}
This work was supported by the Science and Technology Department of Shanxi Province under grant JB/201903D321015.
Keywords: Biodosimetry, cytokinesis-block micronucleus, dose-response curve, X-ray
References
IAEA (International Atomic Energy Agency). Cytogenetic Dosimetry: Applications in Preparedness for and Response to Radiation Emergencies. Vienna: IAEA-EPR; 2011.Fenech M. Nat Protoc 2007;2:1084-104.Diana C, Popescu IA, Cîmpeanu MM. International Conference on e-Health and Bioengineering (EHB). IEEE; 2020. p. 1-4.
Abstract - 43188: Quality assurance programme embraced in personal monitoring services at TLD Laboratory, Kudankulam Nuclear Power Project
K. B. Jashi, A. Ashok Kumar, P. Pandaram
TLD Laboratory, Kudankulam Nuclear Power Project, Kudankulam, Tamil Nadu, India
E-mail: [email protected]
Introduction: Personal monitoring services (PMS) in nuclear power plant play an imperative role in Radiation Safety. Competency of radiation safety to radiation worker could be ensured by accuracy in assessment of individual dose. The demands on the requirement of individual monitoring in terms of accuracy, performance and recording level are high. Diverse internal quality control measures are implemented in Personal Monitoring Services at TLD laboratory, through engineered controls and administrative controls, which facilitates accuracy in estimation of individual dose.
Uncertainties in PMS: In general, the uncertainty of a dosimetric system is determined from the combined effects of the two types of uncertainty namely the Type A, random and Type B, systematic. Type A uncertainties are those which can, in principle, be reduced by increasing the number of measurements. Type B uncertainties are those which cannot be reduced by repeated measurements and the sources generally are Energy dependence, Directional dependence, Non-linearity of the response, fading dependent on ambient temperature and humidity.
Internal Quality Assurance (QA) Programme: QA programme at TLD Lab-KKNPP is more emphasised for qualitative and quantitative assessment of external exposures of the radiation workers. It is established to lessen the Type A and Type B uncertainties associated with PMS and to achieve accuracy and consistency in evaluation of personal external exposure. This depends on the quality of TLD cards and performance of the readers. New TLD cards are tested for initial QA check comprising physical inspection and sensitivity check for acceptability and their performance is evaluated periodically. Readers are checked for the stability of EHT, Heating unit, light source output, dark current, Linearity and reproducibility, residual TL and Glow curve pattern regularly. In addition, purity of nitrogen gas and performance of annealing oven are checked on monthly basis.
Human Error Reduction Techniques Adopted: Human error in the area of calibration is found to be one of the contributors for Type B uncertainty. In addition, there is possibility of human error in other activities such as data entry, inventory check, data analysis and Report generation.
Administrative Controls: It includes implementation of custom made Software for dose estimation, Checklist Cum Datasheet for Internal Quality assurance checks (QAC), Peer Review of Data, Colour Codification Of TLD Wrapper/Cassette, Upkeep of Maintenance history card and Preset datasheet for readout of service cards.
Engineered Controls: Replacement of Perspex calibration table with Customized auto irradiator system facilitates better accuracy in dose delivery and reproducible geometry. Implementation of Multichannel Nitrogen Gas distribution System ensures uninterrupted supply of Nitrogen gas for TLD processing from Nitrogen generators (2 Numbers) as main source and Nitrogen cylinders as backup.
Result and Discussion: The end result of implementing various equality control measures were confirmed from the results of external QA check done by RPAD, BARC and Blind test by user group. The results were subjected to two performance evaluation standards namely ANSI 2009 Criteria and ISO Trumpet curve analysis. [Graph 1] shows the total uncertainty estimated are well below the acceptable Tolerance level of 0.3.[INLINE:6]
Conclusion: An identified uncertainty in Personal Monitoring Services is dealt with systematic and rigorous internal quality assurance programme ensured smooth functioning of PMS at KKNPP. The potential human error possibilities in various stages of PMS are controlled by implementation of appropriate human error reduction techniques through administrative and engineered control measures.
Keywords: ANSI, ISO, personal monitoring services, TLD
References
ANSI. American National Standard for Dosimetry – Personnel Dosimetry Performance- Criteria for Testing. ANSI/HPS N13; 2009.Hand Book on the Use of TLD Badge based on CaSO4: Dy Teflon TLD Discs for Individual Monitoring (BARC/2002/E/025).
Abstract - 43277: Kinetic parameters of CaSO4: Dy teflon embedded disc
P. Pravin Kumar1, K. S. Reddy1, A. K. Bakshi1,2, B. K. Sapra1,2
1Bhabha Atomic Research Centre, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
CaSO4:Dy thermoluminescence (TL) material embedded in Teflon discs are extensively used in India for personal monitoring for mixed fields of γ,β and X-rays owing to its high sensitivity, simple glow curve structure and low fading rates. In the present study, kinetic parameters of the CaSO4(Dy) Phosphor were studied using the glow peak shape method. Samples of the CaSO4:Dy Teflon TL discs of 0.8 mm thick, 13 mm diameter and weight 280 mg were exposed to gamma dose of 80 mSv using 137Cs source with calibrated output traceable to national standard. The matrix contains 70 mg of the TL phosphor and 210 mg of Teflon.[1] Prior to irradiation, TL discs were annealed at 2300C for 4 h to remove any previous TL signal. Read out of the discs were carried out on a PC based TL reader model: TL 1009I procured from M/s Nucleonix, Hyderabad. TL glow curves of 10 discs were acquired with heating of 0.50C/s after 1 week of exposure. TL Kinetic parameters were determined from the shape of the isolated glow peak [Figure 1] using Chen's method.[2] The TL parameters such as geometrical shape factors such as τ, δ, ω, μ and γ were computed based on the glow peak shape methods, {Figure 53}
Where,
τ - Tm-T1 is half width at low temp side of the glow peak.
δ - Tm-T2 is half width toward the fall off-side of glow peak
ω – T2-T1 ids total half width, is called symmetrical factor, is geometrical factor,
k- Boltzmann's constant.
The graphical picture of dependence of the symmetry factor suggested by Chen has been utilized to determine the order of kinetics b. Based on the experimental values of μ= 0.54±0.01 and γ =1.20±0.07, the order of kinetics was determined as 2.10. Trap depth E (eV) was computed using equation-1 for various computed geometrical factors such as τ, δ and, ω and are tabulated in [Table 1]. The results were in agreement with similar studies carried out by Azorin and Guttierez.[3] However it may be noted that kinetic parameters due to temperature lag especially frequency factor is influenced. Hence frequency factor has not been determined using the present methodology. Due to temperature lag, there will be uncertainty in other kinetic parameters. Sample TL Glow curve is shown [Figure 2].{Table 28}{Figure 54}
[INLINE:7]
where α is τ, δ, or ω and the values of cα and bα are summarized below
cτ = 1.510 + 3.0(μ − 0.42), bτ = 1.58 + 4.2(μ − 0.42)
cδ = 0.976 + 7.3(μ − 0.42), bδ = 0
cω = 2.52 + 10.2(μ − 0.42), bω = 1
The activation energies obtained by the Chens method using the parameters μ, δ and ω, E(eV) and order of kinetics b were are in good agreement among them and found to be consistent with literature.
Keywords: CaSO4(Dy), kinetics, thermoluminescence
References
Vohra KG, Bhatt RC, Chandra B, Pradhan AS, Lakshmanan AR, Shastry SS. A personal dosimeter TLD badge based on CaSO4: Dy Teflon TLD discs. Health Phys 1980;38:193-7.Chen R. J Electrochem Soc Solid State Sci 1969;116:1254-7.Azorin J, Guttierez A. Thermo Chim Acta 1988;135:121-5.
Abstract - 43384: Development of software application for quick scan whole body monitor
Puneet Jindal, Rajesh Sankhla, Aatef Shaikh, M. K. Sharma, P. D. Sawant, Probal Chaudhury
Radiation Safety Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
QSWBM is an in-vivo monitoring system that measures internal contamination due to high-energy photon emitters in the human body. In order to automate the monitoring process, a Quick Scan Monitoring (QSM) software application has been designed and developed. QSM application consists of the Server application, Local machine application and Panel PC application. Requests for monitoring are generated by swiping a RFID based identity card or manually by the supervisor. An individual report is automatically generated after each monitoring, and monthly reports are generated periodically by the supervisor.
Application Flow: The monitoring request is raised either automatically by swiping RFID identity card or manually by the Supervisor using the web application. Personal details of the radiation worker are fetched from the Card reader, and the person gets registered automatically if not already registered. After registration, the Local machine provides voice instructions to the radiation worker. On following the voice instructions, weight, height and photograph of the worker are acquired automatically using BMI machine and camera respectively. The local machine uses proximity sensors to detect the worker's presence inside the monitoring chamber. After detection, it starts the spectrum acquisition for preconfigured time. Once the acquisition is over, monitoring data is saved in the web application, and individual report is generated. Based on the spectrum analysis, the Local machine instructs the radiation worker to either go or contact the Supervisor. Facility has been provided to generate various reports which are saved in the Web Application database.
Design Details: Role-based Access Control has been implemented to assign different access permissions to different user groups: “Supervisor” and “Operator“. Supervisor can manage monitoring parameters and generate reports in PDF, “Operator” can perform background acquisition, subject spectrum acquisition, and FWHM (Full-Width Half Maximum) acquisition. Web application and Local machine presents users with login form for authentication before authorizing to perform various tasks. As shown in [Figure 1], web application has three modules: Supervisor module, Monitoring interface and Reporting module. Supervisor module implements functionalities required by “Supervisor” role such as configuring acquisition time, background spectrum, energy calibration, region of interest, dose coefficients, retained fractions etc. Monitoring interface implements functionalities required by “Operator” role such as remote triggering of subject monitoring using Message Queuing Telemetry Transport (MQTT),[1] energy calibration, FWHM and efficiency calibration spectrum acquisition, new worker registration etc. Reporting module is used by Supervisor to generate monthly reports and various quality assurance reports such as background reports, energy calibration reports, FWHM reports etc. The front end of web applications is developed in Vue.js and Bootstrap 4, while the backend is implemented using Web2py[2] framework and Python language. Open source PostgreSQL Relational Database Management System (RDBMS) has been used for data storage. Local machine uses MQTT to send voice messages to the Panel PC. Local machine and Panel PC are multithreaded applications written in C#.NET and WPF. The local machine also uses PostgreSQL database which is the real-time backup of web application database implemented using streaming replication.{Figure 55}
Results and Discussion: QSM application is being used with QSWBM in IDS Lab RSSD at BARC Hospital. It has automated the process of worker registration, whole-body monitoring, spectrum and dose analysis, report generation and maintenance.
Keywords: Access control, automation, database, spectrum, web
References
Available form: https://www.hivemq.com/blog/mqtt-essentials-part-1-introducing-mqtt/2015.Di Pierro M. web2py. Lulu.com; 2013.
Abstract - 43389: Performance in onsite assessment and external quality assurance of TLD Lab, PFBR
G. Shanthi, T. Sudhasini, N. Suriyamurthy, Allu Ananth
Personnel Monitoring Laboratory, BHAVINI, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: BHAVINI, a 500 MW Prototype Fast Breeder Reactor is currently marching towards commissioning. TLD based Personnel Monitoring service for radiation worker is done to cater the regulatory requirements as per the accreditation guidelines provided by RPAD, BARC. This paper outlines the statistical analysis and overall performance of TLD Lab, PFBR achieved during 45th and 46th External Quality Assurance test and Onsite Assessment Proficiency/Spot test conducted by RPAD, BARC for all accredited TLD laboratory.
Materials and Methods: Onsite Assessment (OSA) done during accreditation renewal to check lab infrastructure, maintenance of NODRS, Auto-Irradiator system, Oven, TLD Badge Readers and nitrogen generator required for processing of TLD cards. During OSA by RPAD, BARC spot test conducted to check the dose evaluation technique and they verified results of previous external QAC. Annealed TLD cards were sent to BARC for participating in spot test and external quality assurance check as per the format given by RPAD, BARC for proficiency checking. These TLD cards were exposed to various doses in BARC. Spot test TLD cards were processed in presence of BARC official during onsite assessment visit. External QAC cards were processed and results sent to RPAD, BARC to verify dose evaluation of the TLD Laboratory. Results of 45th External QAC (36 No's), 46th QAC (16 No's) and spot test (15 No's) cards with photon category were subjected for the test. Identification of radiation type and Dose evaluations done for this above test was confirmed based on ANSI criteria and Trumpet Curve Analysis method by RPAD, BARC and performance of PFBR TLD lab is evaluated. In this paper, statistical analysis of 45th, 46th External QAC and spot test dose performance of the TLD cards were taken with photon category radiation. The tolerance level for each test and also overall tolerance is calculated. Trumpet curve for the three different tests done in a single graph [Figure 1]. Number of cards with the performance percentage is arrived.{Figure 56}
Results and Discussion: The bias B, standard deviation S and the tolerance level L is consolidated for the 3 test, performed after accreditation is shown in [Table 1]. As per ANSI criteria, if tolerance level (L) is less than 0.3 the performance is satisfactory for all radiation categories. Overall Bias factor is 0.03, Standard deviation 0.07 and tolerance level is 0.08 which is well within the acceptance limit. Trumpet curve is drawn for the two QAC test and one spot test conducted for photon category in [Figure 1]. It is seen that for all the gamma photons in the plot between true dose and Measured Dose /true Dose, all the value lie within the range (95%) as trumpet curve analysis method. [Table 2] shows the number of cards lying in the different performance coefficient range for photon. It is noted that during the spot test 50% of TLD cards have performance Quotient within +0.05 and 100% of the TLD cards have performance Quotient of +0.10. During QAC test +0.25 performance Quotient is achieved by 100% of the TLD cards for the photon category.{Table 29}{Table 30}
Statistical study shows that QAC and spot test results of TLD Laboratory, PFBR are satisfactory with good number of cards with overall tolerance of 0.08 for photon category. The TLD card quality and sensitivity is maintained to achieve such results.
Reference
American National Standard for Dosimetry of Personal Dosimetry Performance Criteria for Testing ANSI; (ANSI: HPS N13.11-2009); 2009.
Abstract - 43400: Performance and validation of TLD badge reader for emergency dose assessment
T. Sudhasini, G. Shanthi, N. Suriyamurthy, Allu Ananth
Personnel Monitoring Laboratory, BHAVINI, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: TLD Laboratory, PFBR is accredited by RPAD, BARC to accomplish TLD Based-Personnel Monitoring Services (PMS) to plant personnel. TLD Badge Reader plays a key role as one of the major instruments required for TLD Laboratory; its performance is verified as per procedures given in BARC Report.[1] This paper outlines the results for reader response obtained by processing of TLD cards which is exposed to various doses (Dose:1 mSv, 3 mSv, 10 mSv, 50 mSv, 100 mSv, 250 mSv, 500 mSv, 750 mSv, 1 Sv, 1.25 Sv to 1.5 Sv) including emergency exposure level.
Materials and Methods: In TLD Badge Readers (Make: Nucleonix; Model No: TL 1010 S), reader calibration factor (Dose: 3 mSv) on monthly basis and linearity of reader (Dose Ranges: 0.3 mSv to 30 mSv) on annual basis is verified. To validate the response of reader beyond linearity exposure limit and to check for emergency dose level, TLD cards were exposed to 10 different dose ranges in Regional Calibration Facility, IGCAR and processed in TLD Badge Readers. Precautions are taken to avoid fatigue of the PMT during processing of high exposed card.
Results and Discussion: Exposed TLD cards are processed in Reader 1 and Reader 3. Processed data is analyzed as per linearity method, ANSI criteria and trumpet curve analysis.
Test for Linearity: [Table 1] indicates for Reader 1 and Reader 3, relative response (Ri) is within ± 10% for all dose ranges and Coefficient of variation (CoV) is within acceptable limit(Dose ≤ 1.1 mSv : CoV ≤ 15% and Dose> 1.1 mSv : CoV ≤ 6%). Saturation dose of reader is not attained up to 1.5 Sv. It is observed from [Figure 1], for Reader 1 and Reader 3 response is linear for the dose range from 1 mSv to 1.5 Sv.{Table 31}{Figure 57}
ANSI Criteria: Performance of these TLD cards is tested using ANSI Criteria applicable for personnel dosimetry. Based on the processed data, the following ANSI parameters, Performance quotient (Pi), Standard Deviation (S), Tolerance Level (L) are obtained and given in [Table 2]. For individual dose range performance coefficient is obtained. The overall tolerance level (L) is within the accepatable limit (0.30) for both readers.{Table 32}
Trumpet Curve: Trumpet Curve Criteria is adopted to verify the reader response and the results are depicted in [Figure 2]. It is observed from the above graph, in the plot between True Dose to Measured Dose/True Dose the entire dose values are fit within the two limits(Hll and Hul) specified in graphical representation for both readers. Hence the reader performance is satisfied.{Figure 58}
Conclusion: Reader performance is verified with Linearity, ANSI and Trumpet Curve Analysis for various dose levels in Reader 1 and Reader 3. Performance of reader is tested and found satisfactory for emergency dose assessment and both readers are validated to process TLD cards with doses up to 1.5 Sv.
Reference
Datta D, et al. Handbook on TLD-Based Personnel Monitoring (Rev.1-2018). BARC Report; 2018.
Abstract - 43431: Dose evaluation algorithm and routine calibration method for implementation of Hp(10) in personnel monitoring
S. M. Pradhan1,2, Munir S. Pathan1, Shatabdi Chakrabarty1, Minal Punekar1, T. Palani Selvam1,2
1Radiological Physics and Advisory Division, BARC, 2Homi Bhabha National Institute, Mumbai Maharashtra, India
E-mail: [email protected]
TL dosimeters based on CaS04:Dy embedded Teflon discs and hot N2 gas based TLD badge readers are being used for countrywide personnel monitoring of radiation workers in India. The old quantities whole body dose, beta dose / skin dose are still continued for dose evaluation and reporting. These are not consistent with the internationally recommended operational quantity, Hp(d). An algorithm[1] employing linear combination of response of TLD discs was developed to calculate Hp(10). The coefficients of this algorithm are recalculated to account for proper calibration. Subsequently the modified algorithm has been recommended by a task group in RP&AD, BARC for dose evaluation. This paper presents the performance characteristics of the proposed algorithm and evaluates various methods for routine calibration to implement the operational quantity. The algorithm for Hp(10) is as follows:
[INLINE:8]
If ρ≥0.8, Hp(10) = N1 × RCF, where N1, N2 and N3 are net readings of Discs D1, D2 and D3, respectively, N23 = (N2 + N3)/2, ρ = N1/ N23 and RCF is reader calibration factor. The performance of the algorithm was analysed using the data sets of TLD badge response obtained in various studies (algorithms preparation, in-house and EURADOS inter-comparisons). The response to various photon beams of energy 17 keV – 1.25 MeV, angles of incidence 0 - ± 60o, mixture of different photon beams, photon- beta mixtures of 137Cs and 90Sr-90Y / 85Kr and delivered Hp(10) in the range of 0.21 – 440 mSv is included. [Figure 1] shows frequency plot of Q values (ratio of measured Hp(10) to delivered Hp(10)) indicating variation within ± 30% for most of the response data (500 Nos.). The uncertainty in the Q is about 11% at 95% confidence assuming expanded uncertainty 5% and 10% in delivered Hp(10) and measured Hp(10) respectively. There is almost no under-response beyond -30%. However, over-response of about 30-60% to photon energy 60-65keV are observed possibly due to limitation of the fit and badge design. The algorithm is found to be very stable indicating consistently smaller coefficient of variation of the identically exposed badges.{Figure 59}
Routine Calibration: Routine calibration which involves determination of calibration factors for individual detectors or dosimeters or batch of dosimeters can be carried out in free air.[2],[3] Further, radiation field of less than the highest metrological quality, such as beta source for calibration of Hp(10) dosimeter can be used for the type of dosimeter whose characteristics once determined remains constant. For such simplification, it is essential to establish the correspondence between the readings of the dosimeter or detector in the test field and that under type test conditions. This correspondence may be established using samples of dosimeter itself as transfer instruments. In case of Indian TLD badge system, routine calibration involves determination of RCF for the batch of TLD cards and DCF (Disc Calibration Factor) for the reader/cards. In one of the methods, TLD badges (TLD card in TLD cassette) are exposed free in air under panoramic geometry to calibrated 137Cs source. The response under such condition is found to be related with that under type test as: 300 mR =0.96*10* 300 = 2880 μSv of Hp(10). However, DCFs are not evaluated using the cards exposed in cassettes. The other method involves irradiation of TLD cards in 3 mm Perspex build-up under the above setup. Correspondence of response in this case is observed to be 300 mR = 1.03*10*300 = 3090 μSv of Hp(10). This method has an advantage of simultaneous evaluation of DCFs. Routine calibration based on beta source such as 90Sr/90Y built into an irradiator is an attractive option as the use of 137Cs and associated radiation protection issues may be avoided. It is planned to develop and characterise beta irradiator for routine calibration.
Keywords: Algorithm, Hp(10), personnel monitoring, routine calibration
References
Pradhan SM, Sneha C, et al. Radiat Prot Dosimetry 2009;136:176-84.Bartlett DT, Alberts WG. Radiat Prot Dosimetry 1994;54:259-66.IAEA SRS No. 16. Calibration of Radiation Protection Monitoring Instruments. Vienna: IAEA; 2000.
Abstract - 43514: Study on gamma dose mapping using thermoluminescent dosemeter badge in a research reactor
M. Bhattacharya, S. Kasthuri, Lalit Vajpayee1, M. T. Valvi1, A. K. Bakshi, Ranjit Sharma1
Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 1Health Physics Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
A study was conducted for the assessment of gamma dose at various locations inside workplace of a research reactor building. The objective of the study was to compare cumulative dose data between Thermoluminescent Dosemeter (TLD) and Electronic Personal Dosemeter (EPD) during reactor operation at selected locations and correlate with the anticipated personal doses during the monitoring period. The results of the measurement can help to understand the pattern of TLD badge response and also appropriate action in case of unusual patterns and high doses. The different areas of reactor building are classified as Supervised and Controlled areas.[1] The radiation survey was carried out inside reactor building at various locations in controlled areas during reactor operation before commencing the dose assessment. The initial radiation level measurement was conducted by Portable survey meter (RadEye-G Personal Radiation detector). The exposure charts were prepared based on the gamma dose rates to identify the locations of workplace monitoring. The cumulative gamma dose measurements were carried out using (TLD) and (EPD). TLD is personal monitoring badge based on CaSO4:Dy Teflon embedded three Disc D1, Disc D2 and Disc D3 (TLD).[2] and EPD (DMC 3000)[3] is a energy compensated, double silicon detector, and measures individual dose equivalent Hp(10), It covers photon energy range 15 keV – 7 MeV with accuracy ± 30 %. For dose mapping, both types of dosemeters (TLD & EPD) were exposed together for a period of 45 days including 4 days of reactor shutdown. The periodic radiation survey was carried out to compare the data with initial survey results and detection of any fluctuations in the radiation field. All TLDs were processed as per the procedures followed in personnel monitoring[2] and arrived at the cumulative dose in terms of operational quantity H*(10). Gamma dose equivalent rate H*(10) were in range of 0.03 - 60 μSv/h based on radiation survey. Average ambient dose equivalent measured at locations under supervised area for the duration of 45 days using TLD was about 0.05 ± 0.09 mSv. Readout pattern of each TLD deployed at controlled area was analyzed to identify the radiation type and disc ratio of each TLD D2/D1, D3/D1 and D3/D2 indicates the effective energy[2] of radiation type. These disc ratios indicate that most of the pattern related to photon energy >200 keV. Further measurements of both dosemeters (TLD & EPD) were compared and found to agree well. Cumulative dose measured by TLD and EPD dose are presented in [Table 1]. It may be noted that the dose measured by TLD and EPD are in terms of H*(10) and Hp(10) respectively. Doses measured by these dosemeters are comparable.[4] This study provides detail on radiation type, energy, cumulative dose which is useful information in personnel monitoring. The generated data of dose profile expected at different locations in areas of reactor building can be a useful for the investigation of genuineness of high exposures/unusual pattern. It can be inferred that only a fraction of measured cumulative doses may be received by the radiation workers during normal operations due to limited working hours. However, the individual dose depends on occupancy of radiation worker in the area and location.{Table 33}
Keywords: Cumulative gamma dose, electronic personal dosemeter, thermoluminescent dosemeter
References
Radiation Protection Manual for BARC Facilities. BARC/SM/23020/3 REV.-0; 2020.Datta D, et al. Hand Book on TLD – based Personnel Monitoring. 2018. BARC/2018/E/007; 2018.Barthe J, et al. Electronic dosimeters based on the solid state detectors. Nucl Inst Meth Phys Res B 2001;184:158-89.Singh VP, Managanvi SS, Bihari RR, Bhat HR. Operational experience of electronic active personal dosemeter and comparison with CaSo4:Dy TL dosemeter in Indian PHWR. Radiat Prot Dosimetry 2013;156:93-102.
Abstract - 43524: Investigation on dosimetric issues about high-dose inhomogeneous radiation accidents: evaluation of effective dose and organ doses
M. Kowatari, H. Yoshitomi1, Y. Tanimura1, O. Kurihara
Department of Radiation Measurement and Dose Assessment, National Institute of Radiological Sciences, National Institutes for Quantum Science and Technology, 1Nuclear Science Research Institute, Japan Atomic Energy Agency, Tokai, Ibaraki, Japan
E-mail: [email protected]
A nuclear and radiological accident is not frequent. However, there are 634 identified radiological accidents, according to the systematic review of reports from 1980 to 2013.[1] 169 out of 634 cases are overexposure accidents involved in the industrial sector. This study investigated dosimetric issues in high-dose inhomogeneous external exposure which may happen in an industrial radiography. Workers engaged in industrial radiography are often designated as radiation workers in compliance with domestic regulations against radiation protection. Wearing personal protective equipment including personal dosimeters (PDs) for radiation workers is mandatory and doses received should be recorded for the legal purpose. When an exposure accident involving radiation workers takes place, the readings from PDs worn by them will provide useful information in determining the treatment plan for the exposed personnel. In cases where radiation workers are homogeneously exposed throughout the body to radiations i.e., gamma-rays from sealed sources, readings from PDs are an appropriate indicator for medical triage to exposed victims. On the other hand, On the other hand, readings from PDs are not always an indicator of exposure status particularly in inhomogeneous exposure situations. This is because that the PDs worn on the worker's trunk might not be exposed to radiation and the local skin and other organs are likely to be highly exposed. From the viewpoint of the legal issue, there is still room for consideration of how the effective dose should be evaluated in the cases of inhomogeneous exposure accidents. Reporting doses to regulatory authority is required in terms of effective dose. In this study, a Monte Carlo (MC) calculation code was employed to investigate homogeneous and/or inhomogeneous exposure scenarios with reference to actual radiation accidents[2],[3] and to evaluate the relationship between readings from PDs, organ doses, and computed effective doses. A MC code PHITS 3.24 was employed for a series of calculations.[4] The International Commission on Radiological Protection (ICRP) computational adult male phantom was introduced as a radiation worker.[5],[6] As exposure scenarios, in addition to the ICRP-specified irradiation condition of anterior-posterior (AP) irradiation geometry, point radiation source was set on the right chest or on the left hip to mimic inhomogeneous exposure accidents. For these exposure scenario, gamma-ray from 137Cs was chosen. To obtain the readings from PDs, OSL dosimeter consisting of Al2O3 was simulated and set on the left chest of the adult male phantom. [Table 1] summarizes the comparison of calculated doses of radiation worker who might be homogeneously exposed or locally exposed by point 137Cs gamma-ray source. As shown in [Table 1], the PD reading indicated conservatively in AP irradiation (homogeneous exposure). However, PD readings obtained in each inhomogeneous exposure situation were ten times smaller than computed effective doses. Detailed discussion on relationship between computed effective doses and readings from PDs will be made in the presentation.{Table 34}
References
Coerytaux K, et al. PLoS ONE 2015;10:e0118709.IAEA. The Radiological Accident In Gilan, International Atomic Energy Agency, Vienna; 2002.IAEA. The Radiological Accident in Ventanilla, International Atomic Energy Agency, Vienna; 2019.Sato T, et al. J Nucl Sci Technol 2018;55:684-90.ICRP. The 2007 Recommendations of the International Commission on Radiological Protection. ICRP Publication 103. Ann ICRP 2007;37.ICRP. Conversion Coefficients for Radiological Protection Quantities for External Radiation Exposures. ICRP Publication 116. Ann ICRP 2010;40.
Abstract - 43539: Performance of TLD personnel monitoring services provided by TLD lab, IGCAR in external quality assurance tests
O. Annalakshmi1,3, G. V. Bharathilashmi2, V. Subramanian1,3, Shailesh Joshi1, C. V. Srinivas1,3, B. Venkatraman1,3
1Safety Quality and Resource Management Group, Indira Gandhi Centre for Atomic Research, 2Health Physics Division, Bhabha Atomic Research Centre Facilities, Kalpakkam, Tamil Nadu, 3Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
In India, Individual monitoring of all radiation workers is being carried out using the three element Thermoluminescent dosimeters (TLDs) based on CaSO4:Dy embedded Teflon discs. The increase in the usage of radiation sources and increase in the number of radiation workers along with the stringent regulatory requirements had led to the decentralisation of TLD dosimetry services which have led to the establishment of many accredited personnel monitoring services laboratory. The TLD laboratory at IGCAR is one such accredited laboratory and it caters to the dosimetry requirements of all radiation workers of IGCAR and BARC facilities at Kalpakkam. It was set up in the year 2000 and was accredited by BARC in the year 2001. As part of the requirement of the accreditation and in order to have uniformity in the dose evaluation procedures routine external quality assurance (QA) tests along with proficiency tests are being conducted by BARC. Apart from this, the lab also participated in the EURADOS international intercomparison exercises. This paper presents the performance analysis of the laboratory from the results of external QA tests during last 20 years.
External Quality Assurance Tests: Routine periodic QA test conducted by BARC, proficiency tests conducted during renewal of accreditation and EURADOS international intercomparison exercises are the external QA tests participated by the laboratory. This external QA tests checks the overall performance of a monitoring laboratory and it helps to demonstrate and document the uniformity of procedures adopted by different monitoring laboratories. In the case of routine periodic QA and EURADOS intercomparison exercises, annealed TLD cards / badges selected randomly from the service cards will be sent to BARC or the coordinating laboratory for irradiation. After irradiation, the cards will be returned to the monitoring laboratory and the doses will be evaluated using the standard procedures followed for routine cards. Once the dose values are communicated the performance of the laboratory will be evaluated based on ANSI criteria and trumpet curve analysis.
Materials and Methods: Detector: Regular TLD badges used for personnel monitoring in India (CaSO4:Dy embedded Teflon discs) Semi-automatic TLD badge reader (Nucleonix TL 1010S).
Results and Discussion: TLD lab at IGCAR has participated in 22 external QA test conducted by BARC and the results of each QA tests was analysed by trumpet curve and ANSI criteria. The tolerance level (L) of the laboratory for different QA cycles along with the allowed tolerance level used for evaluation is given below in [Figure 1]. The performance of the laboratory was satisfactory in all the external QA cycles. Apart from this the lab has also participated in several proficiency tests and three EURADOS intercomparison exercises. The trumpet curve results of the latest proficiency test, external QA and Hp(10), Hp(0.07) values of EURADOS intercomparison is given in [Figure 2] below. The trumpet curve analysis also shows that the performance of the TLD personnel monitoring laboratory is satisfactory in measuring the radiation dose due to different radiation qualities and quantities.{Figure 60}{Figure 61}
Keywords: Personnel monitoring services, quality assurance, thermoluminescent dosimeters
References
BARC. External Report No. BARC/2018/E/007.Lakshmanan, et al. Radiat Prot Dosim 1989;28:273-6.
Abstract - 43541: Correlation between TLD and DRDs exposed to mixed fields of Cs-137 and Se-75
G. V. Bharathilashmi, E. Yasotha1, I. Vijayalakshmi1, M. G. Komathi1, O. Annalakshmi1,2, C. V. Srinivas1,2
Health Physics Division, Bhabha Atomic Research Centre Facilities, 1SQ&RMG, Indira Gandhi Center for Atomic Research, Kalpakkam, Tamil Nadu, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: As per the atomic energy act 1962 and the associated radiation protection rules 2004, all radiation workers should be individually monitored for both external and internal exposure. External exposure received by radiation workers (from nuclear power plants, reprocessing facilities, medical facilities and industrial radiography) are being monitored using three element thermoluminescent dosimeters (TLDs) based on CaSO4:Dy embedded Teflon discs and the processing of these dosimeters is carried out using semiautomatic TLD badge readers. Normally electronic pocket dosimeters (EPDs) are being used along with the legal dosimeters TLDs for day to day monitoring and control of radiation doses received by radiation workers. TLD–DRD dose discrepancy is being investigated at facilities where both active dosimeters (DRD/EPD) and passive dosimeters like TLDs are being used systematically throughout the monitoring period like at all Nuclear Power Plants sites in India whereas in other facilities like reprocessing facilities, research laboratories handling radiation, radiography workers etc, use both TLDs and DRDs only for some special works taken up through radiological work permit. In this work a systematic study has been undertaken to study the variation between TLD and present DRDs (EPDs).
Materials and Methods: Detector: Regular TLD badges used for personnel monitoring in India (CaSO4:Dy embedded Teflon discs), Semi-automatic TLD badge reader (TL 1010S Nucleonix make) semiconductor based EPDs (Thermoscientific make – 15 keV to 10 MeV), Cs-137 source and Se-75 source for panaromic exposure.
Results and Discussion: In this study, a correlation was established between the TLD doses and EPD doses under laboratory conditions to two different gamma ray sources depicting mixed energies encountered in the field conditions. The experiment was carried out under different conditions like (i) exposing TLDs and DRDs to single source (ii) Exposing to mixture of radiation (both Cs-137 with energy 662 keV and Se-75 with average energy 0.371 MeV) (iii) Exposing to fractional dose (in steps) upto a cumulative dose of 1 mSv and 3 mSv using both the sources depicting field conditions. A good correlation was observed between the TLD doses and DRD doses when they are exposed to either Cs-137 or Se 75 or combination of both the radiations or in the case of fractionated dose experiments. Also the variation of the TLD dose and EPD dose with respect to the true dose is shown in [Figure 1]. Under laboratory conditions, the percentage variation between the doses measured by the TLDs or EPDs is within ±20% only for the combination of both irradiations. However the variation between true dose and dose recorded by EPD varied upto around 40% in the low dose region (0.1 mSv) which is the recording level and is within ±10% at dose levels above 1 mSv, whereas with TLDs the variation is within ±20% from the true dose at all dose levels. Trumpet curve analysis of both the dosimeters was carried out for all the dose levels and exposure situations. It is found that all the data points are within the limits of trumpet curve which further confirms that the discrepancy in either the TLD dose / EPD dose from the true dose is within the acceptable limits of dosimetry standards under all conditions. This study suggests that with the present EPDs, the discrepancy between TLD and EPDs will be within the acceptable limits prescribed by the task force report on TLD-DRD discrepancy.{Figure 62}
Keywords: EPDs, personnel monitoring services, thermoluminescent dosimeters
Reference
Handbook on TLD-Based Personnel Monitoring (Rev. 1-2018), BARC/2018/E/007.
Abstract - 43552: Variation in photon response of 3-element extremity and eye-lens dosemeter on rod and cylindrical phantom
Kshama Srivastava, Shatabdi Chakrabarty, Rohit, S. K. Singh1
Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, 1Radiation Standards and Systems Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Monitoring of extremity and eye-lens doses of radiation workers is required to ensure compliance with the dose limits recommended to protect these organs from tissue reactions. The extremity and eye-lens doses dosemeters are used in non-uniform radiation fields and workplaces with potential for significant exposure to these organs than whole body e.g. interventional radiology, nuclear medicine, radioisotope production etc. A new compact size 3-element Extremity Ring Badge (ERB) and Eye lens Badge (ELB) dosemeters have been developed at BARC for monitoring extremity and eye lens dose of radiation workers in India.[1],[2] These dosemeters are based on the indigenously developed CaSO4:Dy Teflon discs(0.4 mm thick, 5 mm dia). The dosemeter comprises 3 detector discs placed under suitable filters inside a plastic holder and provided with Velcro™ based size adjustable band/strap [Figure 1]. The mass thickness of 3 filters above Disc 1, Disc 2 and Disc 3 are 7 mg/cm2, ~1000 mg/cm2 and ~300 mg/cm2, respectively. The ERB and ELB dosemeters have same detector-filter pack, however the calibration method and dose evaluation algorithms are different. The response of these two dosemeters measured on different phantoms is compared in this paper. The pre-selected TL Discs within ± 5 % sensitivity range were used after annealing at 350°C for 1 h. The ISO 4037-1 recommended Narrow Series X-ray beams (N 15, N40, N60, N80, N100, N150, N200 N250) and S-Cs Gamma were used in the study.[3] The beam output was provided by RSS, RSSD. The ERBs were placed on surface of ISO Rod Phantom (1.9 cm dia, 30 cm height) and ELBs on ISO Cylindrical Phantom (20 cm dia, 20 cm height) for irradiations at 2 m for X-rays and 1 m for 37Cs gamma. TL research readers (RPIS, BARC and Intech) were used for readout at 280°C (35 s, 10°C/s). The ERB and ELB dosemeters are calibrated in terms of Hp(0.07) and Hp(3) on Rod and Cylindrical Phantom respectively, for 137Cs gamma.[3] Based on the photon energy/angular response, a suitable algorithm was developed to estimate photon dose equivalent for both dosimeters as below:{Figure 63}
ERB dose = Hp (0.07) = N1 x C. F. (Photon, Rod) (1)
ELB dose = Hp (3) = N3 x C. F. (Photon, Cyl.) (2)
Where N1 and N3 are net disc reading under Mylar and Teflon filter, respectively and C.F. is correction factor to compensate energy/angular response of the detector. The ERB and ELB response was evaluated and compared for various photon fields on Rod and Cylindrical phantom. [Table 1] gives ratio of Disc ratio R12 (R12ELB,Cyl/ R12ERB, Rod) and Disc 1 response(N1/Hp(0.07)Rod / N1/Hp(3)Cyl,) and ISO conversion coefficient (hp,K(0.07)Rod / hp,K(3)Cyl) at normal incidence for extremity/eye lens dosemeters. The effect of phantom size seems less prominent on disc response ratio than Disc Ratio as delivered dose is accounted for change in phantom size/configuration. The ERB and ELB dosemeters are capable to measure the dose due to gamma, beta and low energy X-ray radiations and discriminate the beam quality as well.{Table 35}
Keywords: CaSO4:Dy teflon disc, extremity dose, eye lens dose, personnel monitoring, thermoluminescence
References
Srivastava K, et al. Abstract Book of IARPNC-2014; 2014. p. 155.Srivastava K, et al. AMPICON 2019, Kolkata, India, November 7-9, 2019, ABS 0153.ISO 4037-3:2019.
Abstract - 44154: Evaluation of Burlin general cavity theory in the kilovoltage region
Sudhir Kumar, S. D. Sharma
Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Cavity theory is fundamental for understanding and predicting the detector response. When a significant proportion of the absorbed dose in a detector arises from electrons generated by photon interactions in the detector material and a proportion arises from electrons generated in the medium, the cavity is termed as 'intermediate-size cavity'.[1] Burlin general cavity theory (BGCT) was developed to deal with these intermediate cavities in photon beams.[2] According to this theory, the dose to the water at a D(z)w specified position, z can therefore be written.
[INLINE:9]
where [INSIDE:1] is the Spencer-Attix, mass stopping-power ratio, water-to-det, for a cut-off energy Δ, and D̄det is the average dose over the sensitive volume of the detector. The [INSIDE:2] is the ratio of the mass energy-absorption coefficient of the water to the det, 'd' is the weighting factor which gives the contribution to the total dose of medium (water) generated electrons and (1-d) is the contribution to the total dose from electrons generated by photon interaction in the detector/cavity. Mobit et al.[1] used Monte-Carlo (MC) technique to examine the BGCT and obtained more precise values of the 'd' for several TLDs materials irradiated by megavoltage photons in PMMA, water, aluminium, copper and lead phantoms and also quantified the accuracy of the BGCT against MC estimates of the dose ratio. But they have not examined/evaluated the accuracy of BGCT in kilovoltage (kV) region. In this paper BGCT was examined for x-ray beams of 50, 100, 150 and 250 kV for a small air cavity on the central axis in water medium.
Materials and Methods: The sensitive volume of the real NACP-02 plane-parallel ionization chamber was modelled as a cylinder with thickness (0.2 cm) and cross-sectional area (radius 0.5 cm). The detector voxel comprising a single volume of sensitive material (air), surrounded by liquid water, was located at 2 g cm-2 depth along the central axis of the beam in a cylindrical water phantom (15 cm radius, 30 cm thickness) with SSD of 25 cm. The dose to air was computed using the EGSnrc user-code CAVRZnrc[3] for beam radius of 5 cm defined on the phantom surface for 'clinical' kV spectra (for point sources) at 50, 100, 150 and 250 kV. The 'Source 1' option of the EGSnrc MC code system (i.e. point source, incident on front face) was used. The dose to water was also scored in the identical dimensions/equisized voxel for the scoring volume (0.5 cm radius, 0.2 cm thickness). Thereafter the dose ratio water-to-air was computed. The calculations were carried out with the default settings. A PEGS4 datafile was generated with EGSnrcMP package with parameters AP = 1 keV, AE = 512 keV (total energy) where AP and AE are the production thresholds for secondary bremsstrahlung photons and knock-on electrons respectively. Electrons and positrons were followed down to 1 keV kinetic energy (i.e. the electron/positron kinetic energy cut-off ECUT = 512 keV) and photons down to 1 keV (photon energy cut-off PCUT = 1 keV). The [INSIDE:3] (with Δ = 10 keV) were computed using user-code SPRRZnrc for beam radius of 5 cm defined on the phantom surface for 'clinical' kV spectra mentioned above. The [INSIDE:4] were also computed using MC simulations for above mentioned geometry and 'clinical' kV spectra. By selecting the parameter PHOTON REGERATION = 'no electrons from wall' (i.e. IFANO = 2), in CAVRZnrc user-code, the 'd' and (1-d) for air cavity were calculated. Thereafter, the equation (1) was used to determine the dose ratio water-to-air.
Results and Conclusions: [Table 1] presents the dose ratio water-to-air computed by MC simulation and estimated by BGCT using all parameters computed using MC method. It is seen from [Table 1] that [INSIDE:5] is never greater than 77.2% for any of the kV x-ray qualities being investigated. This is a clear demonstration that air-filled cavity placed in water phantom do not fulfil Bragg-Gray conditions in kV x-ray beams. The ratio of dose-to-water to dose-to-air computed by MC, [INSIDE:6] and determined by BGCT, [INSIDE:7] is in good agreement (within 2.8%) for all kV beam qualities used in this work. Burlin general cavity theory was found to work well for a small air cavity on the central axis in a water phantom{Table 36}
Keywords: Absorbed dose, Burlin general cavity theory, EGSnrc Monte-Carlo, kilovoltage photons
References
Mobit, et al. Phys Med Biol 1997;42:1319-34.Burlin TE. Br J Radiol 1996;39:727-34.Rogers, et al. NRCC Report No. PIRS-702 (rev C). Ottawa: National Research Council of Canada; 2019.
Abstract - 44298: Simulation based estimation of dosimetric quantities for ICRP computational and CIRS phantoms due to galactic cosmic rays at LEO
Rohit, Sandipan Dawn, A. K. Bakshi, B. K. Sapra
RP&AD, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
During space missions in LEO or ISS, there is a high concern of exposure to the astronauts due to space radiation in which galactic cosmic rays (GCR) plays an important role. Generally different types of phantom are used to find out the effect of these radiation to the critical organs of human being during space missions. Ideally anthropomorphic phantom or human torso phantom as recommended by ICRU should be used. In practice, phantoms with materials having electron density and effective atomic no different from the one recommended by ICRU, may lead to equivalent dose to organs different from the ideal one. In this Project, Monte Carlo Code FLUKA (FLUktuierende KAskade)[1] was used to compare the dose, dose equivalent and average quality factor (<Q>) for several organs of Atom Phantoms made by CIRS (Computerized Imaging Reference Systems) with respect to the reference man organ materials as per ICRP 110.[2] For solar modulation Φ = 465 MV (solar minimum), particles from z = 1 to z = 28 were transported for energy range from 1 MeV/amu to 103 GeV/amu as shown in [Figure 1], as this part mainly contribute to dose and dose equivalent. Particles were transported isotropically from spherical source in inward direction where initially ICRU sphere was placed as target. USERBIN card of FLUKA was used to calculate absorbed dose and dose equivalent in target materials due to GCR spectrum. ECUT off for all particles is 100 keV. Up to 107 particle histories were followed in each simulation and statistical uncertainty was less than 2%. Absorbed dose and dose equivalent in ICRU sphere was calculated as 0.40 mGy/day and 1.47 mSv/day respectively which are comparable with results of Ballarini et al.[3] After the validation of results, a sphere of radius 10 cm was taken as target organ having varying compositions from CIRS and ICRP 110 for different organs to compare dosimetric quantities. Details of simulated dose, dose equivalent and <Q> for different organs are shown in [Table 1]. Although the composition of CIRS and ICRP materials were different, however, the absorbed dose and dose equivalent values per day and average Q factor for both type of phantoms were found to agree well. In most cases, <Q> simulated were around 4.1, which indicate that the dose delivered was mainly due to the low LET charged particles. The study indicates that, CIRS phantoms can be used to estimate the dosimetric quantities for human organs during space missions using passive detectors.{Figure 64}{Table 37}
References
Battistoni G, Boehlen T. Overview of the FLUKA code. Ann Nucl Energy 2015;82:10-8.ICRP. ICRP Publication 110 Adult Reference Computational Phantoms. ICRP; 2009Ballarini F, Battistoni G, Cerutti F, Fasso A, Ferrari A, Gadioli E, et al. GCR and SPE organ doses in deep space with different shielding: Monte Carlo simulations based on the FLUKA code coupled to anthropomorphic phantoms. Adv Space Res 2006;37:1791-7.
Abstract - 44337: Estimation of theoretical efficiency of LR-115
Dibyendu Rana1, V. N. Jha1, R. L. Patnaik1, M. K. Singh1, S. K. Jha1,2, M. S. Kulkarni1,3
1Health Physics Division, Bhabha Atomic Research Centre, Departments of 2Chemical Sciences and 3Physical Sciences, Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: Ionizing radiations within suitable energy range and incident angle are capable of producing tracks in Solid State Nuclear Track Detector (SSNTD) films. The tracks can be microscopically visualized following chemical etching in suitable medium under specified conditions. Using the film radon (222Rn) dosimeters has been developed and is in use for uranium mine environment of Jha.[1] Tracks formed are converted into radon exposure using experimentally determined calibration factor and efficiency. Present study provides numerically estimated calibration factor of LR-115 film inside a prototype cylindrical diffusion chamber. The findings corroborate with the experimentally obtained calibration factor estimated by Jha.[1]
Materials and Methods: LR-115 film with dimension 4cm X 2.25 cm is mounted at the bottom face of a cylinder having radius r=2.95cm and height H=3.5 cm. The top face of the cylinder is covered by a permeable membrane of thickness L=50 μm with diffusion coefficient D=6.34 X 10-4 cm-2.s-1. For large exposure time, the short lived radon daughters (Po218, Pb214, Bi214) will be in equilibrium with the radon inside the cylinder. If Cin in Bq m-3 is radon concentration inside and Cout in Bq.m-3 is radon concentration outside the chamber, they can be related as equation,[1] [INSIDE:8]
Where λ is radioactive decay constant of radon =2.1 X 10-6 s-1.
At equilibrium, all the short lived radon daughters should have identical activity = Cin V, which is distributed at the surface due to deposition and volume of the cylinder. Deposited activity fraction (fi) of the daughters is listed in [Table 1].[1] {Table 38}
If fi (i=1,2,3 for Po218, Pb214, Bi214 respectively ) is the deposited fraction for ith daughter nuclei, then surface activity is, [INSIDE:9] In Bq m-2
In equation (2), V is volume of the cylinder in m3 and S is total surface area inside cylinder in m2. Similarly, the volume fraction for ith daughter is,
[INSIDE:10] in Bq.m-3
Due to non interactive nature, Rn222 do not deposit on the surface and will remain inside the volume with concentration as shown in equation (1) Rn222 along with its daughters emits alpha particle in energy range ~ 6 - 8 MeV. Upon chemical etching (60 0C, 10% NaOH, 2 h), only charged particle within the energy range 0 - 4 MeV will produce visible tracks.[1] The distances between which each radio nuclei can produce the visible tracks are tabulated in [Table 2].[1]{Table 39}
All the radio nuclides presents within R1 and R2 from LR-115 will produce tracks for incident angle less than 400.[3] Using these assumptions, the visible tracks can be estimated by,
[INLINE:10]
from surface deposition and contribution from volume deposition is given by,
[INLINE:11]
In above equations R1≤R≤R2 which are different for different nuclei along with incident angle ≤400 will give visible tracks.
Conversion factor can be defined as
[INLINE:12]
Here s' represent surface area of the LR-115 film.
Results and Conclusion: Using equation (6) the estimated calibration factor is E = 0.549 T which is in agreement with the experimentally obtained result 0.4852 T.[1] Here E is in Bq.l-1.h and T is in Tracks cm-2. The generalized approach for calibration factor estimation through measuring the surface deposition fraction can be extended for all the progeny at different wall under variable atmospheric conditions.
Keywords: Radon progeny, solid angle, tracks
References
Jha G. Development of a Passive Radon Dosimeter for Application in Radiation Protection and Uranium Exploration. India: University of Bombay, PhD Thesis; 1986.Markovic VM, et al. RN progeny diffusion, deposition and track distribution in diffusion chamber with permeable membrane. Radiat Meas 2019;124:46-157.
Abstract - 44475: Effect of lung size on the response of HPGe detector based lung monitor
M. Manohari, Kevin Capello1, R. Mathiyarasu, D. Ponraju, B. Venkatraman
Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, India, 1Human Monitoring Laboratory, Radiation Protection Bureau, Ontario, Canada
E-mail [email protected]
In IGCAR, coaxial HPGe detector of diameter 8.5 cm and thickness 3.03 cm having a Be entrance window of thickness 0.08 mm placed tangential to the sternal notch is used for the estimation of actinides in lungs. This detector is calibrated using Lawrence Livermore National Laboratory (LLNL) realistic phantom having active lung set of 21 cm length with uniform distribution of radionuclides. In case of gamma energy less than 100 keV the scattering factor due to the variation in the dimension of organ size is 1.5. In this paper the effect of the lung size variation for this geometry with detector to phantom distance of 1 cm is studied numerically using two LLNL voxel phantoms having lungs of different dimensions but same volume for a chest wall thickness of 1.6 cm. Kramer and Capello,[1] has reported that for a lung monitor having an array of four coaxial HPGe detectors each of 7 cm diameter the counting efficiency of 241Am varied by a factor of 1.44 for a variation of 10 cm in the length of the lungs keeping the volume unchanged. As the energy increased the difference reduced. In case of low energy (<100 keV) the effect of different components of the detector such as dead layer, protective grove, bulletization etc., are significantly high on the efficiency. Numerical model of the detector with all the above components was constructed using the FLUKA code [Figure 1] along with the LLNL phantoms [Figure 2] having lungs of 21 cm (a) and 11 cm (b) length but of same volume. The phantom was modeled using 3D square lattice structure. Efficiencies were obtained for different energies using DETECT card. FLUKA model of the detector and the phantom were validated using experimental efficiency of LLNL phantom having 241Am loaded long lung set. The deviation between the measured and simulated efficiency is 0.3%. [Table 1] compares the efficiencies of different radionuclides for both the lung sizes.{Figure 65}{Figure 66}{Table 40}
The difference in efficiency varied by a factor of 1.9 (239Pu), 1.6 (241Am), 1.4 (137Cs). So, one has to have an exact replica of the subject being counted to reduce the error in the dose estimation. Compared to Kramer and Capello,[1] in the current geometry the deviation is more due to lesser active area as well as the location of the detector. The results of this study show that the effect of lung size on the efficiency is more predominant when the area covered by the counting system is small. Optimized surface area and position of the lung monitor would reduce the error further with respect to lung size.
Keywords: Efficiency, HPGe detector, LLNL voxel phantom, lung size
Reference
Kramer GH, Capello K. Lung Counting: Effect of Lung Volume on the Counting Efficiency of a Four-Detector Array Using 70 mm Detectors. HMLTD-03-10; 2016.
Abstract - 45139: Beach sand quartz as a retrospective dosimeter
A. Ashok Kumar, M. G. Komathi1, O. Annalakshmi1, B. Sundarakannan
Department of Physics, Manonmaniam Sundaranar University, Tirunelveli, 1Health Safety and Environment Group, Indira Gandhi Center for Atomic Research, Kalpakkam, Tamil Nadu, India
E-mail: [email protected]
Introduction: Quartz, being a radiosensitive material, its thermoluminescence (TL) dosimetric properties is being investigated at large so that it can be used as a fortuitous dosimeter in the event of any emergencies. When quartz extracted from beach sand has been investigated for dosimetric application, the major difficulty was to quantify the fading component because beach sand will be exposed to sunlight during the day time and dark environment in the night times. The fading effect of the peaks of the quartz is due to the charge carriers escaping from the trapping centers within the forbidden gap in quartz Jose et al.[1]
Materials and Methods: Quartz was extracted from beach sand without the application of heat using mechanical methods and by using the difference in the specific gravity of heavy minerals and quartz using bromoform. Single aliquot mode was utilized for the study. For each measurement 10 mg of the sample was used. The dry Quartz (125-micron grain size) sample was then exposed to a dose of 50 Gy using Co-60 gamma source in a gamma chamber and the fading studies were carried out. The samples were read using Riso TL/OSL-DA-20 reader. The low temperature peak was identified at 88°C but was not considered for the studies due to its rapid fading rate. Apart from the low temperature peak two peaks were identified at 203 and 325°C. at a heating rate of 5°C/sec. The peaks are labels as Peak-1 and Peak-2 respectively. Both the peaks were obtained after optimum pre heat of 155 and 260°C respectively.
Results and Discussion: It was observed that the rate of fading in dark environment in one hour is -6% and -3% of the initial intensity for peak-1 and peak-2 respectively. The fading factor was estimated to be 4.254 x 10-3 and 9.137 x 10-4 for peak-1 and peak-2 respectively in the dark environment at room temperature. The fading factor of the samples kept under sunlight was found to be 2.322 x 10-2 and 2.139 x 10-2 for peak-1 and peak-2 respectively. The samples kept under sunlight showed a fading of -13% and -30% in the initial one hour but later it was found that the fading of Peak-1 is higher compared to peak-2. About 64-70 % of the TL intensity is lost within one week post exposure. Commercial glass showed 3-5%, and 35–71% fading under dark and sunlight storage conditions, respectively Pradeep et al. (2007).[2] It can be inferred that even in dark environment the fading happens due to the phosphorescence and room temperature. Under sunlight the UV rays, visible and IR rays are the contributing factors for fading.
Conclusion: Hence with the knowledge of the fading trend with time and the time elapsed between exposure and measurements to carry out, quartz extracted from beach sand can be used for retrospective evaluation of dose.{Figure 67}{Table 41}{Table 42}
Keywords: Fading, quartz, retrospective dosimeter
References
Jose MT, Anishia SR, Annalakshmi O, Ramasamy V. Radiat Meas 2011;46:1026-32.Pradeep N, Senwar KR, Vaijapurkar SG, Kumar D, Bhatnagar PK. Appl Radiat Isot 2008;66:86-9.Aydaş C, Aydin T. Appl Radiat Isot 2015;101:65-74.
Abstract - 45190: Measurement of annual dose using thin layer of LCAF phosphor
Sonal Kadam, S. N. Menon, Bhushan Dhabekar
Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected]
Introduction: The determination of annual dose is one of the most important part of retrospective dosimetry. For the internal component of the annual dose in soil/pottery it is sufficient to determine only the beta component. Beta dosimetry using TL technique has been used widely for the determination of annual dose.[1],[2] LiCaAlF6:Eu,Y (LCAF) is the indigenously developed high sensitive OSL phosphor with 0.5 μGy minimum detectable limit[3] (as compared to MDL of ~5 μGy of Al2O3:C). High sensitivity of LCAF phosphor makes it possible to determine the dose in shorter time duration as compared to commercial available Al2O3:C. The possible use of a thin layer of LCAF phosphor for the measurement of beta dose from various soil samples was studied. Estimated beta doses were compared with the dose levels obtained from standard samples and from the samples of known concentration of Uranium (U), Thorium (Th) and Potassium (K).
Materials and Methods: LCAF phosphor in powder form was used for the beta dosimetry. The OSL measurements were performed using an automatic Risø TL/OSL reader (TLDA20) at the stimulation power of 72mW/cm2. A 90Sr/90Y β source with a dose rate of 20 mGy/s available in the reader was used for delivering the calibration doses. Different soils of known dose-rates (standard samples) and soils with known concentrations of U, TH and K were used for the comparison of estimated doses. The Perspex sample holder of 0.8cm depth and 2.5cm diameter were fabricated for the experiment. The sample holders were then filled with the soil samples and covered with a thin Mylar foil to prevent the alpha radiation from reaching the phosphor. Thin layer of LCAF phosphor was spread uniformly on Mylar sheet. The sample holders were then kept in a lead container for sufficient period of time and the OSL from phosphor was recorded. Doses were also obtained by Monte Carlo (MC) simulations using FLUKA general purpose MC code. Geometry of sample (i.e. 0.8 cm depth and 2.5 cm diameter), description of soil matrix (whole container filled with dry soil without water content), type of source (Volume source) and OSL phosphor (radiation detector) were used as Input parameters.
Results and Discussion: Doses were estimated using recorded OSL from LCAF phosphor by applying appropriated calibration factor. These measured doses were compared with doses calculated using the known concentration of U, Th and K in the soil sample and is shown in the [Table 1]. The estimated doses using thin layer of LCAF phosphor were matching well with the calculated doses.{Table 43}
Keywords: Annual dose, beta dose, LCaAlF6:Eu,Y, OSL, retrospective dosimetry
References
Goksu HY, Bulur E, Wahl W. Radiat Prot Dosimetry 1999; 84:451-5.Bailiff IK, Aitken MJ. Nucl Instrum Methods 1980;173:423-9.Dhabekar B, Rawat NS, Gaikwad N, Kadam S, Koul DK. Radiat Meas 2017;107:7-13.
Abstract - 45286: Can non-destructive electron paramagnetic resonance tooth dosimetry be used for posterior assessment of radiation exposure in medicine and dentistry?
I. Yamaguchi 1, Y. Nakai2, M. Miyake1,2
1Department of Environmental Health, National Institute of Public Health, Wako, 2Department of Oral and Maxillofacial Surgery, Kagawa University, Miki, Japan
E-mail: [email protected]
Background and Purpose: Assessing radiation exposure among patients and health care workers is a challenge issue. Electron paramagnetic resonance (EPR) tooth dosimetry is an established method for evaluating the absorbed dose ionizing radiation by measuring the unpaired electrons to in-teeth samples. Most of the previous studies regarding EPR tooth dosimetry required the removal of the material with processing in vitro. To carry out radiation dosimetry in live subjects such as for potential radiation emergencies, a lower-frequency L-band EPR apparatus that can be applied to teeth in situ has been developed for measurement under triage conditions [Figure 1]. This method was also applied to some animals affected by the Fukushima Daiichi Nuclear Power Station accident which occurred in March 2011.[1] In the present study, we demonstrate the possibilities of implementation of non-destructive EPR tooth dosimetry for medical radiation exposures.{Figure 68}
Methods: This EPR tooth dosimeter operates in continuous-wave mode and uses at an excitation frequency near 1.15 GHz with 0.4 mT modulation at 20 kHz. We measured human teeth before and after irradiation using a 4 mT sweep range. Each sweep of 3 seconds was repeated 20 times. The detection limit of this method was compared to the doses to patients and health care workers from radiation exposure in medicine.
Results and Discussions: A detection limit of 2 Gy was achieved for gamma rays due to Cs-137 with in vivo EPR tooth dosimetry. The detection limit for X-rays in the diagnostic area was 0.5 Gy because the sensitivity was about four times higher than gamma rays due to Cs-137. The median free air kerma per dental intraoral radiograph was estimated to be 3.6 mGy;[2] 140 radiographs would reach 0.5 Gy, so 7 radiographs per year would reach a detectable level in 20 years [Figure 2]. For dental cone-beam CT with a FOV of 40 to 100 cm2, the air kerma at the center of the beam in a single examination is 29 mGy,[3] and the detection limit level can be reached in 17 examinations. Medical workers engaged in endoscopes are more likely to have an equivalent dose to the lens of eye of more than 20 mSv/ year.[4] The detection limit is reached in 10 years for medical workers with an equivalent dose of 50 mSv/y in the lens of the eye when protective eyewear is not considered. The differential absorption of radio waves by unpaired electrons produced by radiation is shown. The signal on the right is due to the positive control.{Figure 69}
Conclusions: This method is also applicable for retrospective dose assessment not only for a large-scale radiation exposure incident, but also for radiation exposure in medical exposure both patients and workers.
Keywords: Dental radiology, electron paramagnetic resonance, occupational radiation exposure, patient, tooth enamel
This work was supported by the KAKEN under grant 18K09724.
References
Yamaguchi I. et al. Appl Sci 2021;11:1187.Sakaino R. et al. Shika Hoshasen 2010;50:51-7.J-RIME. Japan DRL 2020; 2020.IAEA. SSG-46; 2018.
Abstract - 45438: Electron paramagnetic resonance fingernail dosimetry for rapid individual dose assessment
Madhusmita Panda1, Shailesh Joshi1, O. Annalakshmi1,2, C. Venkata Srinivas1,2, B. Venkatraman1,2
1Environmental Assessment Division, Indira Gandhi Centre for Atomic Research, Kalpakkam, Tamil Nadu, 2Homi Bhabha National Institute, Mumbai, Maharashtra, India
E-mail: [email protected]
Widespread use of ionizing radiation in many fields like industry, agriculture, medicine, research etc. has increased the probability of unintentional exposure of occupational workers and public as a result of radiation accidents. Hence it is needed to develop techniques for retrospective dosimetry providing rapid and accurate dose estimation, particularly useful for individuals not equipped with physical dosimeters. For dose reconstruction in this scenario, methods usually based on free radicals are used. Electron Paramagnetic Resonance (EPR) dosimetry is such a method based on measurement of stable radiation induced free radicals in biomaterials like tooth enamel, bones, fingernails, etc.[1] in the human body. EPR tooth enamel and bone dosimetry cannot be easily applied in the early phase of the accident because of the invasiveness of the sampling. Hence EPR dosimetry on fingernails becomes more attractive. The scope of the present work was to investigate the EPR dosimetric characteristics of fingernails and develop a methodology for rapid dose assessment to individuals. Fingernail samples used in this work were collected from adult volunteers (male and female; Indian type). Prior to experiments, nail samples were soaked in distilled water for 10 min and dried in air for 15 min in order to eliminate the initial clipping effect during sample collection and to remove the dirt.[2] The EPR measurements were performed using a Bruker EMX X-band EPR spectrometer with a high Q resonator. Gamma irradiations were performed at room temperature using 60Co source calibrated in terms of air kerma. [Figure 1] represents the EPR spectra of unirradiated and irradiated (5 and 10 Gy) fingernails recorded immediately after cutting. From the figure it can be observed that the low field minimum of the signal is mainly due to irradiation (dose dependent) and the high field minimum is due to mechanical cutting (dose independent). This spectral distinction can be used to rapidly screen the population in case of radiation emergencies. For the quantitative estimation of dose during rapid screening in order to seek proper medical guidance, a dose response curve based on the relative intensities of the low field minimum and high field minimum of irradiated fingernails recorded immediately after cutting has been established and shown in [Figure 2]. The response was found to be linear up to 30 Gy. This study was carried out on freshly exposed fingernails and also on irradiated nails stored for three days. This method of establishing the dose response curve for dose estimation is designated as rapid screening method for the first time. From the fading study of fingernails, it was observed that signal faded to 50% within 7 days and remained stable up to 30 days. The suitability of the proposed method was checked by comparing with other reported methods of dose estimation and the result was found to be satisfactory. Also, this methodology was validated with different standard dosimeters like TL discs and Alanine/EPR and the result showed that the estimated doses were within ±15% of dose estimated from the standard dosimeters. The results of this study suggest that the new methodology can be adopted for rapid assessment of dose during radiation emergencies.{Figure 70}{Figure 71}
Keywords: Dosimetry, electron paramagnetic resonance, fingernails, rapid screening
References
Wieser A, Haskell E, Kenner G, Bruenger F. EPR dosimetry of bone gains accuracy by isolation of calcified tissue. Appl Radiat Isot 1994;45:525-6.Wilcox DE, He X, Gui J, Ruuge AE, Li H, Williams BB, et al. Dosimetry based on EPR spectral analysis of fingernail clippings. Health Phys 2010;98:309-17.
Abstract - 45600: Deciphering impression of high LET radiation (14 MeV DT-neutron and Am-α) on lymphocyte chromosome: Application of multiplex and multi-BAND-FISH
Rajesh K. Chaurasia1,2, K. B. Sirsath1,2, N. N. Bhat1,2, B. K. Sapra1,2
1Homi Bhabha National Institute, 2Radiological Physics and Advisory Division, Bhabha Atomic Research Centre, Mumbai, Maharashtra, India
E-mail: [email protected], [email protected]
Introduction: Biological effects of radiation is the function of LET (linear energy transfer) i.e., energy deposition pattern of the radiation (photon/particulate). Quantifiable structural-chromosomal changes induced by low and high LET radiation are not alike.[1] Till now, biodosimetry, clinical impact and risk estimates of exposure to high LET radiation is not very clear.[2] The identification and establishment of a signature marker for high LET radiation and its dose response curve for different LET of radiation, their track structure and the resulting biological effects are still under investigation/exploration. This study intended to identify signature structural chromosomal changes induced by high LET radiation and their correlation with doses, for instance 14 MeV DT-neutron induced intrachromosomal changes were measured, in the lymphocytes of 3 volunteers using advance technique, multi-BAND-FISH.
Objective: This study aims to characterize structural and numerical chromosomal changes, as a function of LET of radiation, using 14 MeV DT neutrons, 3.5 MeV Am-α and 60Co-g-ray. Also, aiming to establish dose response-correlation with specific/signature chromosomal changes.
Methodology: Human lymphocytes were obtained by density gradient centrifugation and irradiated with various doses of 14 MeV DT neutron, 3.5 MeV Am-α and 60Co- g-rays. Irradiated lymphocytes were cultured and metaphases were harvested. Chromosomes were hybridized with advance multiplex and multi-BAND-FISH probes, to observe and quantify, fine structural details, with the help of automated AXIO imager Z2 microscope, installed with ISIS software.
Results and Discussion: Multi-BAND-FISH, offered great insight visualization of complex chromosomal changes with high precision and great improvement in the limit of resolutions. 14 MeV DT-neutron and 3.5 MeV α irradiated metaphases were analysed and it was observed that intrachromosomal changes were predominantly induced and can be considered as an indicator of high LET radiation exposures. To generate the dose-response correlation, 14 MeV DT-neutrons and 3.5 MeV α induced, intrachromosomal (interstitial deletions [Figure 1]a plus intrachromosomal rearrangements) and complex rearrangements [Figure 1]b were quantified in the blood sample of 3 volunteers (age, 27, 36 and 42 years, all male), in the dose range of 0-200 mGy. Independent (for complex rearrangements) and pooled (for interstitial deletion + intrachromosomal rearrangements) dose-response curves were generated [Figure 1]d and [Figure 1]e. Data fitting showed a linear dose-response relationship with the slope, 0.0015 ± 0.00006 and 0.00429 ± 0.00017 interstitial deletions + intrachromosomal rearrangements/cell/Gy for neutron and α particle respectively. 0.000662 ± 0.0000554 and 0.00154 ± 0.00006 complex intrachromosomal rearrangements on exposure of 1 Gy DT-neutrons and Am-α respectively. Yield of intrachromosomal changes were higher (~ 2.8- 3.1 times) in lymphocytes irradiated with equal dose of Am-α over DT- neutron. The uncertainties in the measurements can be attributed to manual band counting. To minimize this, images captured in each filter set were analysed independent and were matched with superimposed one and the pattern of band available in software. Conclusion: In conclusion, interstitial-deletions, intrachromosomal rearrangements and complex rearrangements were major intrachromosomal changes observed, induced by high LET radiation (DT-neutrons and Am-α). Among all the events observed, frequencies of interstitial deletions were highest, complex rearrangements were lowest and intrachromosomal rearrangements were in-between. Yields of all aberrations were higher for Am-α over DT-neutrons for same dose points. These established dose-response curves, can be employed for dose estimation of suspected over exposed (with 14 MeV DT-neutrons and Am-α) individuals.{Figure 72}
Keywords: 14 MeV neutron, 3.5 MeV α, intra-chromosomal changes, mBAND-FISH, multiplex-FISH
References
Dugan LC, Bedford JS. Are chromosomal instabilities induced by exposure of cultured normal human cells to low-or high-LET radiation? Radiat Res 2003;159:301-11.Shahmohammadi Beni M, Hau TC, Krstic D, Nikezic D, Yu KN. Monte Carlo studies on neutron interactions in radiobiological experiments. PLoS One 2017;12:e0181281.
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