CURRENT CAPABILITIES OF THE IRD-CNEN WHOLE BODY COUNTER FOR IN VIVO MONITORING OF INTERNALLY DEPOSITED RADIONUCLIDES IN HUMAN BODY

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1 International Joint Conference RADIO 2014 Gramado, RS, Brazil, Augustl 26-29, 2014 SOCIEDADE BRASILEIRA DE PROTEÇÃO RADIOLÓGICA - SBPR CURRENT CAPABILITIES OF THE IRD-CNEN WHOLE BODY COUNTER FOR IN VIVO MONITORING OF INTERNALLY DEPOSITED RADIONUCLIDES IN HUMAN BODY B.M. Dantas, A.L.A. Dantas and E.A. Lucena Laboratório de Monitoração In Vivo Divisão de Dosimetria Instituto de Radioproteção e Dosimetria Comissão Nacional de Energia Nuclear Av. Salvador Allende Rio de Janeiro RJ, CEP , Brasil bmdantas@ird.gov.br, adantas@ird.gov.br, eder@ird.gov.br ABSTRACT Occupational exposure to radioactive materials may occur as a result of a variety of professional human activities, such as in nuclear industry; use of unsealed sources in nuclear medicine, biological research and agriculture; production of radiopharmaceuticals, as well as in mining and milling of minerals associated with naturally occurring radioactive materials. The IRD whole-body counter (UCCI) consists of a shielded room with internal dimensions of 2.5 m x 2.5 m x 2.5 m. The walls are made of steel and have a graded-z interior lining made of 3 mm of lead, 1.5 mm of cadmium and 0.5 mm of copper. Such thin layers are aimed to reduce environmental sources of natural background radiation that would affect the measurements of radionuclides emitting low energy photons. An array of four HPGe detectors was used to perform low-energy measurements of radionuclides emitting photons in the energy range from 10 to 200 kev in the lungs, liver and bone tissue. Additionally, one NaI(Tl)8 x4 and one NaI(Tl)3 x3 scintillation detectors are used for measurements in the energy range from 100 up to 3000 kev. A configuration of detector supports allows setting up flexible counting geometries, i.e., whole body and specific organs such as head, lungs, liver and thyroid of an individual laid on a monitoring chair. The UCCI is able to perform in vivo measurement of a large variety of radionuclides emitting photons in the energy range from 10 to 3000 kev. The minimum detectable activities for most of the radionuclides of interest allow its application for occupational monitoring as well as in the case of accidental incorporations. 1. INTRODUCTION Occupational exposure to radioactive materials may occur as a result of a variety of professional human activities, such as in (i) nuclear industry; (ii) use of unsealed sources in nuclear medicine, biological research and agriculture; (iii) production of radiopharmaceuticals, as well as in (vi) mining and milling of minerals associated with naturally occurring radioactive materials. The use of individual monitoring techniques is necessary to control the risks and keep the doses as low and reasonably achievable [1]. The control of internal exposures requires the use of methodologies aimed to identify and quantify a variety of radionuclides in the human body. Such measurements are carried out as part of Monitoring Programmes designed by the Radiation Safety Officer of the Installation and approved by the Nuclear Regulatory Board of the country. In Brazil, the evaluation of Radiation Protection Programmes submitted by Installations is a responsibility of the National Nuclear Energy Commission.

2 The results of in vivo monitoring are also useful for the estimation of the severity of accidents involving incorporation of radioactive materials. In any case, after determining the activity in the body, in Bq, and assuming an incorporation scenario, and applying a dosimetric and biokinetic model of the radionuclide of interest, it is possible to estimate the committed effective dose, in msv [2]. Therefore, the objective of this work is to present the methodologies applied at the In vivo Monitoring Laboratory for the direct measurement of radionuclides in organs and tissues of the human body. 2. MATERIALS AND METHODS One of the main aspects in the design of a whole-body-counter is the shielding, which is responsible for the reduction of the background level generated by the radiation emitted by radionuclides from uranium and thorium natural series, potassium-40, as well as due to the interaction of cosmic rays with the atmosphere and with building materials. The IRD wholebody counter (WBC) consists of a shielded room with internal dimensions of 2.5 m x 2.5 m x 2.5 m. The walls of the room are made of steel and have a graded-z interior lining made of 3 mm of lead, 1.5 mm of cadmium and 0.5 mm of copper. Such thin layers are aimed to reduce environmental sources of natural background radiation that would interfere with the measurements of radionuclides emitting low energy photons. The shielding reduces background in two orders of magnitude in the energy range of the photons of 662 kev emitted by 137 Cs [3]. An array of four high-resolution High Purity Germanium Detectors (HPGe) are used to perform low-energy measurements of radionuclides emitting photons in the energy range from 10 to 200 kev in the lungs, liver and bone tissue. Additionally, one NaI(Tl)8 x4 and one NaI(Tl)3 x3 scintillation detectors are used for measurements in the energy range from 100 up to 3000 kev, in the whole body and in specific organs such as lungs, liver and thyroid. The supports of the detectors allow their positioning over various parts of the body of an individual laid on a comfortable monitoring chair. A series of electronic modules process de signals generated by the detectors, producing a gamma spectrum, which is analyzed, allowing identification and quantification of the radionuclides present in the body of the individual at the moment of the measurement. Table 1 presents a list of radionuclides and corresponding in vivo measurement geometries applied routinely in the IRD whole body counter. It is important to point out that the quality of the measurement and consequently the reliability of the results relies on the calibration procedure adopted. For energies above 100 kev it is a common sense among the operators of whole body counter facilities throughout the world that acceptable calibration factors can be obtained with phantoms produced with polyethylene bottles containing standard solutions of known volumes and traceable activities of the radionuclides of interest. On the other hand, for lower energies, it is advisable to calibrate the detectors with physical anthropomorphic phantoms produced with materials that closely simulate the interaction and scattering of the photons with living matter, i.e., materials presenting attenuation coefficients as close as possible to the living tissues [4].

3 Table 1. List of measurements executed at the IRD whole body counter Radionuclide Geometry Detector 241 Am Head HPGe 241 Am Lungs HPGe 18 F Whole Body NaI(Tl)8 x4 18 F Head NaI(Tl)3 x3 131 I Thyroid NaI(Tl)3 x3 123 I Thyroid HPGe 210 Pb Head HPGe 210 Pb Knee HPGe 210 Pb Lungs HPGe 226 Ra Lungs HPGe 232 Th Lungs NaI(Tl)8 x4 238 U e 235 U Lungs HPGe 1 F&A Whole Body NaI(Tl)8 x4 1 Fission and Activation Products The phantom of the organ of interest containing a known activity of the radionuclide is measured in the standardized geometry. Similarly, an inert phantom is measured to estimate the background count rate and so to determine the net count rate of the phantom. The calibration factor, in cpm/bq, for that specific geometry and radionuclide is calculated as the ration between the net count rate and the activity of the phantom, as follows: CF = ((T-B)/Tc)/Act (1) where CF is the calibration factor, in cpm/bq; Act is the phantom activity, in Bq; T is the Total counts produced by the active phantom; B is the total counts of background; and Tc is the count time, in minutes. The term (T-B)/Tc is called the Net Count Rate of the phantom. The activity of an exposed individual is calculated by dividing the net count rate registered on the in vivo measurement by the calibration factor, as shown in the following formula: Act = cpm/ CF (2) In the case of a lung measurement of a radionuclide emitting photons of low energy it is also necessary to correct the count rate due to the attenuation related to the chest thickness of the tissue over the lung region [5]. Such correction is carried out according to the following expression: CWT = 0,1105 x (w/h) - 2,0038 (3) where CWT is the chest wall thickness, in millimeters; h is the high of the individual, in meters; and w is the weight, in kilograms. The total uncertainty of the in vivo measurement is associated to the uncertainty of the Calibration Factor and the counting of the individual. The main parameters that affect the uncertainty of the measurement are (i) reproducibility of the geometry, (ii) calibration and (iii) distribution of the activity within the organ or tissue [6]. The uncertainties associated to

4 the various parameters of the calculation are propagated according to the following general expression in order to obtain the standard deviation of the activity present in the organ or tissue of the individual: u 2 u =( x 2 2 u ) x +( y 2 2 u ) y +( z 2 2 ) z (4) where u = u(x,y,z), and represents the derived value of the uncertainty. The minimum detectable activity (MDA) is should be calculated for each radionuclide and counting geometry, and is a directly proportional to the square root of the background [7] as shown in the following formula: 4.65 N MDA CF T 3 CF T (5) where MDA is the minimum detectable activity, in Bq; N is the total counts of the background of a unexposed individual in the region of interest; CF is the calibration Factor, in (cpm/bq), and T is the count time, in minutes. The estimation of the committed effective dose for the most relevant radionuclides is carried out by using biokinetic and dosimetric models developed by the International Commission on Radiological Protection (ICRP) and suggested in the publications released by the International Atomic Energy Agency (IAEA). Such models are edited in the data base of the software AIDE [8] allowing the interpretation of the in vivo measurements and the comparison of the internal doses with the limits established by the Brazilian Regulatory Board (CNEN) [9]. It should be pointed out that ICRP classifies the compounds into three categories according to their solubility in the pulmonary fluids: Type F (very soluble), Type M (moderately soluble), and Type S (insoluble). Table 2 presents the parameters used in this paper for the calculation of doses applied to selected radionuclides, considering single intake via inhalation of particles with AMAD equal to 1 m, which corresponds to an inhalable particle size. In the specific cases of 131 I and 123 I it was assumed the intake of volatile elemental iodine, class F. In this work, dose calculations were done assuming an interval of 24 hours between incorporation and in vivo monitoring. Based on such scenarios, retention fractions, in Bq/Bq, and dose coefficients, in msv/bq, were calculated for each organ of interest using the software AIDE. The retention fraction corresponds to the activity present in the organ of interest, after a time t, as a result of the incorporation of 1 Bq. The establishment of the time t, elapsed between the incorporation and the measurement is a critical step in the process of data interpretation, since the parameter m(t) changes as a function of time, according to the dynamic biodistribution of the compound in the body, as well as the radioactive decay of the radionuclide. Both parameters (biological and radioactive half-lives) are expressed in a combined parameter called effective half-life of the radionuclide compound in the body.

5 Table 2. Activities of selected radionuclides in organs of interest, 1 day after the intake of 1 Bq via inhalation (Values calculated using software AIDE version 6) Nuclide Organ of interest Scenario 1 Type/ 2 AMAD Dose Coefficient (msv.bq -1 ) Activity (Bq) 241 Am Lungs M / x x Am Head M / x x Cs Whole Body M / x x F Whole Body F / x x F Head F / x x I Thyroid F, Gas 1.98 x x I Thyroid F, Gas 2.13 x x Pb Head M / x x Pb Knee M / x x Pb Lungs M / x x Ra Lungs M / x x Th Lungs M / x x U Lungs M / x x U Lungs M / x x Solubility class according to ICRP 78 2 Activity Median Aerodynamic Diameter 3. RESULTS AND DISCUSSION Tables 3 presents the calibration results of the HPGe and NaI(Tl) detection systems available at the IRD whole body counter, for a variety of radionuclides. Table 4 presents an evaluation of the sensitivity of the detection system based on a comparison of the minimum detectable activities with the corresponding expected activities in the organs or tissues of interest associated to the respective committed effective doses of 1 and 20 msv. The calibration factors obtained for the various counting geometries present uncertainties in the range of 0.5 to 7%, which is compatible with values usually found in the literature related to in vivo measurement procedures [1]. Standard deviations associated to the calibration factors in cases when estimation is based on calibration curves (americium, uranium and lead in lungs and F&A products in the whole body) is calculated as a propagation of counting statistics and standard source activity uncertainty, and is below 2% for all cases. The use of the software AIDE allows the calculation of the activities, in Bq, which would correspond to the monitoring of an individual 24 hours after the incorporation of the respective radionuclides and would impart committed effective doses of 1 and 20 msv. It should be highlighted that MDA values for the radionuclides and geometries evaluated in this paper refer to the application of the in vivo monitoring technique in situations where it is assumed a single intake, usually associated to incidental incorporation.

6 Table 3. Calibration Factors for the radionuclides monitored in the IRD whole body counter Nuclide Geometry (organ) CF (cpm/bq) 241 Am Lungs e ( CWT) 241 Am Head (skull) F Whole Body (skeleton) F Head (brain) I Thyroid I Thyroid , Pb Head (skull) Pb Knee Pb Lungs e (-0.04.CWT) F&A Whole Body e ( E) 226 Ra Lungs e ( CWT) 232 Th Lungs U Lungs e ( CWT) 235 U Lungs e ( CWT) F&A = Fission and Activation Products CWT = Chest wall thickness (mm) E = Energy (kev) Table 4. Comparison between MDA of the IRD whole body counter detection system and expected activities in measured organ 24 hours after an intake corresponding to committed effective doses of 1 and 20 msv Nuclide Geometry (organ) MDA Activity in measured organ (Bq) (Bq) 1 msv 20 msv 241 Am Lungs Am Head (skull) x x F Whole Body (skeleton) F Head (brain) I Thyroid x x I Thyroid x x Pb Head (skull) Pb Knee Pb Lungs Cs Whole Body x x Ra Lungs Th Lungs U Lungs U Lungs A comparative analysis between the results for head and knee geometries, for the radionuclides 241 Am and 210 Pb, shows that those methodologies are not suitable for their application in a measurement performed 24 hours after the intake. This is in accordance with the biodistribution of those nuclides and their dose coefficients, since the corresponding retention fractions in bone tissue is quite reduced one day after intake [2].

7 On the other hand, in the case of 18 F, considering its low dose coefficient, the in vivo measurement techniques can be considered suitable for monitoring purposes even 24 hours after the intake, since the incorporated activity corresponding to an effective dose of 1 msv is approximately two orders of magnitude higher than the minimum detectable activity. This means that the in vivo measurement technique is suitable for occupational monitoring of 18 F but not for 241 Am and 210 Pb, if the measurement is performed only 24 hours after the intake. In the cases of 214 Am and 210 Pb, an alternative it to perform the measurement after sufficient time has been elapsed so that the expected activities in the organs of interest would reach values compatible with the minimum detectable activities of the detection systems. In other cases like the measurement of 131 I and 123 I in the thyroid, 210 Pb and 235 U in the lungs, and 137 Cs in the whole body, the techniques are comfortably sensitive for their application in the evaluation of accidental incorporations. The techniques are also suitable for the detection of 226 Ra, 232 Th and 238 U at the 20 msv level, considering the incorporation of those radionuclides without their respective progeny. It should be highlighted that the evaluation of each technique in terms of sensitivity for application in routine monitoring requires the calculation of the activity in the compartment of interest considering the time interval between each monitoring. Therefore, the interval between each monitoring should be established so that the minimum detectable activity of the detection system is below the expected activity in the compartment of interest, considering the most likely exposure scenario. 4. CONCLUSIONS The IRD whole body counter is able to perform in vivo measurement of a large variety of radionuclides emitting photons in the energy range from 10 to 3000 kev. The minimum detectable activities for most of the radionuclides of interest allow its application for occupational monitoring as well as in the case of accidental incorporations. References 1. International Atomic Energy Agency (IAEA). Assessment of Occupational Exposure Due to Intakes of Radionuclides - Safety Standards Series, no. RS-G-1.2, IAEA (1999) 2. International Commission On Radiological Protection (ICRP) Publication 78: Individual Monitoring for Internal Exposure of Workers (1998) 3. Oliveira C.A.N., Lourenço M.C., Dantas B.M., Lucena E.A., & Laurer G.R, The IRD/CNEN whole body counter: Background and calibration results, Radiation Protection Dosimetry, 29(3), pp (1989) 4. International Atomic Energy Agency (IAEA). Direct methods for measuring radionuclides in the human body Safety Series 114, Vienna: IAEA (1996)

8 5. D.A., Krushten, & L., Anderson, Improved ultrasonic measurement techniques applied to assay of Pu and other transuranics in lung, Health Physics, 59(1), pp (1990) 6. A., Brodsky, Accuracy and Determination Limits for Bioassay Measurements in Radiation Protection Washington D.C.: US Nuclear Regulatory Commission. Report NUREG-1156 (1986) 7. Health Physics Society (HPS), Performance Criteria for Radiobioassay, HPS N13.30 (1996) 8. Bertelli L, Melo D.R., Lipsztein J., Cruz-Suarez R., AIDE: Internal Dosimetry Software, Radiation Protection Dosimetry, 130(3), pp (2008) 9. Comissão Nacional de Energia Nuclear (CNEN). Diretrizes Básicas de Proteção Radiológica, Norma CNEN-NN-3.01 (2011)

Sensitivity of the IRD whole-body counter for in vivo measurements in the case of accidental intakes

Sensitivity of the IRD whole-body counter for in vivo measurements in the case of accidental intakes B.M. Dantas, E.A. Lucena and A.L.A. Dantas Laboratório de Monitoração In Vivo Divisão de Dosimetria