Applied Radiation and Isotopes 100 (2015) 70–74
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Internal dosimetry of nuclear medicine workers through the analysis of 131I in aerosols Luana Gomes Carneiro n, Eder Augusto de Lucena, Camilla da Silva Sampaio, Ana Letícia Almeida Dantas, Wanderson Oliveira Sousa, Maristela Souza Santos, Bernardo Maranhão Dantas Instituto de Radioproteção e Dosimetria (IRD-CNEN), Av. Salvador Allende s/n, 22783-127 Rio de Janeiro, Brasil
H I G H L I G H T S
The analysis of aerosols in the workplace is recognized as a useful means for controlling and optimizing workers exposure involved in the handling of 131I. Volatilization of 131I occurs during handling of the solutions. Improved ventilation conditions would certainly reduce risks of incorporation.
art ic l e i nf o
a b s t r a c t
Article history: Received 31 July 2014 Received in revised form 12 November 2014 Accepted 24 November 2014 Available online 3 December 2014
131
Keywords: Iodine-131 Aerosols Nuclear medicine
I is widely used in nuclear medicine for diagnostic and therapy of thyroid diseases. Depending of workplace safety conditions, routine handling of this radionuclide may result in a significant risk of exposure of the workers subject to chronic intake by inhalation of aerosols. A previous study including in vivo and in vitro measurements performed recently among nuclear medicine personnel in Brazil showed the occurrence of 131I incorporation by workers involved in the handling of solutions used for radioiodine therapy. The present work describes the development, optimization and application of a methodology to collect and analyze aerosol samples aiming to assess internal doses based on the activity of 131I present in a radiopharmacy laboratory. Portable samplers were positioned at one meter distant from the place where non-sealed liquid sources of 131I are handled. Samples were collected over 1 h using high-efficiency filters containing activated carbon and analyzed by gamma spectrometry with a high-purity germanium detection system. Results have shown that, although a fume hood is available in the laboratory, 131I in the form of vapor was detected in the workplace. The average activity concentration was found to be of 7.4 Bq/m3. This value is about three orders of magnitude below the Derived Air Concentration (DAC) of 8.4 kBq/m3. Assuming that the worker is exposed by inhalation of iodine vapor during 1 h, 131I concentration detected corresponds to an intake of 3.6 Bq which results in a committed effective dose of 7.13 10 5 mSv. These results show that the radiopharmacy laboratory evaluated is safe in terms of internal exposure of the workers. However it is recommended that the presence of 131I should be periodically re-assessed since it may increase individual effective doses. It should also be pointed out that the results obtained so far reflect a survey carried out in a specific workplace. Thus, it is suggested to apply the methodology developed in this work to other nuclear medicine services where significant activities of 131I are routinely handled as an effective means to optimize individual exposures and improve occupational radiation protection safety. & 2014 Elsevier Ltd. All rights reserved.
1. Introduction 131
I is one of the most used radionuclides in Nuclear Medicine due to its high affinity to the thyroid gland tissue, energy and type of n
Corresponding author. E-mail address: [email protected] (L.G. Carneiro).
http://dx.doi.org/10.1016/j.apradiso.2014.11.021 0969-8043/& 2014 Elsevier Ltd. All rights reserved.
radiation emitted. It is used in several procedures including diagnosis and treatment of diseases associated with thyroid. The handling of solutions containing radioiodine can result in a significant risk of occupational exposure due to the possibility of incorporation via inhalation or ingestion (IAEA, 1999; Schwartz et al., 1991). Several studies have been published about occupational exposure in nuclear medicine procedures, where increased use of
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
radioisotopes and availability of suitable methodologies for individual dose assessment have been discussed (Baechler et al., 2011; Bento et al., 2012; Dantas et al., 2007; Dumeau et al., 2011; Heller, 1996; Jansen et al., 1994; Kletter et al., 1999; Vidal et al., 2007). A previous study including in vivo and in vitro methods performed by Lucena et al. (2007), showed the occurrence of 131I incorporation by workers involved in handling of solutions used for radioiodine therapy. The International Atomic Energy Agency recommends, in its Safety Guide Series No RS-G-1.2 (IAEA, 1999), to implement occupational monitoring where the risk of annual effective dose resulting from intakes of radionuclides exceeds the value of 1 mSv. Such document suggests evaluating the workplace through the analysis of aerosols since inhalation is one of the main routes of entry of radionuclides in the human body and the main route of occupational exposure by 131I. This indirect measurement method have a considerable advantage compared to others, since the worker do not have to move from the workplace. Furthermore, it is possible to analyze a great number of samples simultaneously. Due to Brazilian continental dimensions, direct methods of monitoring (in vivo bioassay) is unfeasible if all measurements have to be performed at the institutes of the National Commission of Nuclear Energy (CNEN). In this case aerosol analysis should be the most suitable method for evaluation of these workplaces. In general, elements found in vapor form are not collected efficiently with conventional filters since porosity is not sufficient for its retention. In the case of iodine, specific sorbent materials are used to retain the element. Several research groups have tested methods to collect airborne 131I and activated charcoal was found to be one of the most efficient materials to adsorb iodine because the process is favored by the intrinsic nature of the substrate, i.e., its high porosity and high surface resulting area of the charcoal. Ramos et al. (2013) tested the efficiency of activated carbon for the adsorption of I2, the most common form of volatile iodine found in nuclear medicine, obtaining an efficiency of 100% retention of this chemical form. According to Sales (1981), the use of activated charcoal as a filter medium should be controlled due to its low flash point, so, the smaller the amount of charcoal the greater safety. The analysis of aerosols in the workplace is recognized as a useful means for controlling and optimizing workers exposure involved in the handling of 131I since direct methods are not usually available in all internal dosimetry laboratories. Thus, the aim of this work is to develop a method using activated carbon filters to assess the presence of 131I in aerosol samples collected in Nuclear Medicine Services as a subside to assess internal doses associated to the incorporation of this radionuclide by workers.
2.1. Preparation of filters and experimental arrangement Radiopharmaceutical laboratories usually have reduced dimensions and thus handling of radioactive materials requires high level of attention and care by the professional. Therefore, experimental arrangement was planned in order to cause the least possible inconvenience to routine work. The collection was carried out using a SKC portable Sampler Airchek model 224-44XR and a 37 mm diameter holder with filters containing activated carbon prepared in the Laboratory of Aerosol Characterization of the IRD. The filters, so-called “activated charcoal sandwich”, were prepared using the following materials: 37 mm diameter glass fiber filters; activated charcoal with a granulometry o100 μ and white glue. For each filter, 0.15 mg of charcoal was weighed on a glass fiber filter, and sealed in a sandwich form using white glue applied on the upper edge of the filter. After drying, the ready-made filters are placed in a holder coupled in the suction pump at low flow. Fig. 1 shows the Samplers Airchek and holder. A schematic representation of the samplers positioning is shown in Figs. 2 and 3. It consisted in four samplers, two at the right and two at the left side of the fume hood where 131I solution is handled. 2.2. Determination of
131
I in activated carbon filters
The filter samples containing 131I were analyzed at the In Vitro Bioassays Laboratory of IRD by gamma spectrometry with a Canberras HPGe coaxial detection system operated with the software Genies-2000. The detection system was calibrated with a 133Ba standard source certified by the Ionizing Radiation Metrology National Laboratory of IRD.
2. Materials and methods This study was performed in the Nuclear Medicine Service of a public Hospital located in the city of Rio de Janeiro, whose workers agreed to participate voluntarily in this research. A variety of radiopharmaceuticals are handled in this service, including 131I, the main interest of this work, in the form of capsules and liquid solutions. Since the manipulation of capsules is believed to represent low risk of internal exposure, the investigation was restricted to the radiopharmacy laboratory where iodine solutions are fractionated and doses are prepared to be administered to the patients. A total of sixteen samples were collected during four sampling procedures. The samples were analyzed by gamma spectrometry with Canberras HPGe coaxial semiconductor detector available at the In Vitro Bioassay Laboratory of the Dosimetry Division of the Institute of Radiation Protection and Dosimetry (IRD).
71
Fig. 1. SKC Sampler Airchek model 224-44XR and holder.
72
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
Fig. 2. Schematic representation of the samplers positioning related to the fume hood where
A counting geometry so-called “sandwich” was established in which 0.15 g of an activated carbon was spiked with the 133Ba standard source by gravimetric transference from a plastic vial and sealed between two glass fiber filters (Mcfarland, 1991). The solution was added to the substrate following a geometric model based on hexagonal geometric scheme proposed by Ceccatelli et al. (2010). The activity present in the source at the time of the measurements was corrected according to Eq. (1): − λt
A = A0 x e
(1)
The 131I equivalent activity of the source was calculated according to Eq. (2):
131
I=
131
I solutions are handled by workers.
A131Ba × ∑ γ 133Ba ∑ γ 131I
(2)
The calibration factor was obtained according to Eq. (3):
Fc =
cpm (filter) AEq
(3)
where cpm is the net count rate (count rate in the region of interest subtracted from background count rate), AEq is the 131I equivalent activity of the filter. The calibration of the detection system allows determining 131I activities and concentrations in filter samples.
Fig. 3. Sampling setup: Air filter and pump positioning are planned in order to cause minimum interference on routine work.
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
Table 1 Results for airborne samplings in a radiopharmacy laboratory of a nuclear medicine service in Rio de Janeiro. Sample
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 Average
Collection point
Right (outer face fume hood) Left (outer face fume hood) Right (1 m) Left (1 m) Right (1 m) Right (outer face fume hood) Left (outer face fume hood) Left (1 m) Right (1 m) Right (outer face fume hood) Left (outer face fume hood) Left (1 m) Right (1 m) Right (outer face fume hood) Left (outer face fume hood) Left (1 m) concentration
Filter activity (Bq)
Concentration (Bq/m3)
7.5E þ00 1.7E þ00 2.3E þ00 2.3E þ00 4.5E 03 2.5E 02 7.5E 03 7.4E 03 1.33E 02 3.22E 03 1.20E 02 3.91E 03 o MDA o MDA o MDA o MDA 3 Bq/m3
1.9E þ 01 4.4E þ00 5.8E þ00 5.9E þ00 3.0E 02 1.4E 01 4.5E 02 5.3E 02 3.4E 02 8.2E 03 2.6E 02 7.7E 03 o MDA o MDA o MDA o MDA
n
Minimum Detectable Activity (MDA) ¼ 1.60E 02 Bq.
73
defined as that concentration of airborne activity which would result in the intake of Ij,inh,L by a worker exposed continuously for one year (taken to be 2000 working hours), being Iinhl,L an intake corresponding to the relevant limit on effective dose, e.g., 20 mSv (IAEA, 1999). For a standard breathing rate of 1.2 m3/h, the DAC would thus be given by Eq. (4):
DAC =
I j, inh, L (4)
2000 × 1.2
Applying specific parameters related to 131I in vapor form, DAC is calculated as 416.67 Bq/m³. The experimental data obtained in this work indicate that the exposure estimated over one year of work, for different concentrations are all below the values of DAC. However, it should be taken into consideration that these workers are involved in the handling of other radiopharmaceuticals in the same service, and that this exposure is chronic. It is therefore justifiable to implement actions aiming to reduce 131I incorporations and consequently individual occupational exposures and radiological risks.
3. Results The collection of airborne samples performed at the Radiopharmacy Laboratory investigated in this work used activated carbon filters and an operation flow of 1 L/min. The results obtained are presented in Table 1. The DAC value for 131I in vapor form of 416.67 Bq/m³ is calculated considering a routine workload of 8 h per day and 5 days per week, which results in 2000 h per year of constant exposure over the whole period. However the workers involved in the handling of radiopharmaceuticals do not comply with such regime. Actually the staff perform the tasks in an alternate basis in order to minimize handling and consequently to reduce individual exposure. Based on this actual differentiated work routine, DAC values for different exposure times were calculated considering the manipulation taking place twice a week, and a handling time registered by the workers according to their real work routine. Table 2 presents the parameters used to estimate concentrations under such conditions, and considering the mean activity values for DAC.
4. Discussion The concentration of radionuclides in air is evaluated by the Derived Air Concentration (DAC, expressed in Bq/m3) which is
Table 2 Derived air concentrations of
5. Conclusions The procedures developed in this work and applied for sample collection and radiometric analysis were suitable to the objective of the investigation, both in terms of application easiness and sensitivity required for an assessment of occupational exposure. Based on the results obtained it can be concluded that volatilization of 131I occurs during handling of the solutions. Incorporation was experimentally confirmed through additional monitoring of thyroid and urine of workers involved in the handling, improving sensitivity and applicability of the method to this group of workers. The calculation of the DAC for different working conditions suggests the efficacy of reducing exposure time. Furthermore, improved ventilation conditions would certainly reduce risks of incorporation. Additional studies are needed to verify whether the presence of 131 I in the air is associated with good practices, laboratory design or solely physical and chemical properties of the radionuclide itself. Such information would guide specific recommendations on the improvement of radiological protection conditions of the workplace.
131
I as a function of the time of exposure.
Exposure time (min)
Exposure time (h)
Exposure time/week (h)
Exposure time/year (h)
DAC (Bq/m3)
Concentration (Bq/m3)
10 20 30 40 50 60 70 80 90 100 110 120
0.17 0.33 0.50 0.67 0.83 1.00 1.17 1.33 1.50 1.67 1.83 2.00
0.33 0.67 1.00 1.33 1.67 2.00 2.33 2.67 3.00 3.33 3.67 4.00
17.33 34.67 52.00 69.33 86.67 104.00 121.33 138.67 156.00 173.33 190.67 208.00
4.86E þ04 2.43E þ04 1.62E þ 04 1.21E þ 04 9.71E þ 03 8.09E þ03 6.94E þ03 6.07Eþ03 5.40E þ03 4.86E þ03 4.41E þ03 4.05E þ03
8.67Eþ 00 1.73E þ 01 2.60E þ 01 3.47E þ 01 4.33Eþ 01 5.20E þ 01 6.07Eþ 01 6.93E þ 01 7.80E þ01 8.67Eþ 01 9.53E þ 01 1.04E þ02
74
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
Acknowledgments The authors acknowledge the CNPq for the financial support (Grant number 503563/2012-9). We especially thank the Director of the Hospital, the Head of the Division of Nuclear Medicine and the professionals who participated voluntarily in this research.
References Baechler, S., et al., 2011. Individual monitoring of internal exposure for nuclear medicine workers in Switzerland. Radiat. Prot. Dosim. 144 (1–4), 464–467. Bento, J., et al., 2012. Study of nuclear medicine practices in Portugal from an internal dosimetry perspective. Radiat. Prot. Dosim. 149, 438–443. Ceccatelli, A., et al., 2010. Development of simulated air filters for gamma-ray spectrometry proficiency testing. Appl. Radiat. Isot. 68, 1240–1246. Dantas, B.M., et al., 2007. A protocol for the calibration of gamma cameras to estimate internal contamination in emergency situations. Radiat. Prot. Dosim. 127, 253–257. Dumeau J. et al., 2011. Evaluation de la contamination atmospherique dans les services de medecine nucleaire. IRSN. Heller, S.L., 1996. Radiation safety in the central radiopharmacy. Semin. Nucl. Med. 26 (2), 107–118.
International Atomic Energy Agency (IAEA), 1999. Assessment of Occupational Exposure Due to Intakes of Radionuclides. Safety Standards Series, no. RS-G-1.2, IAEA. Jansen, S.E., et al., 1994. Staff radiation-doses during 8 years in a nuclear-medicine radiopharmacy. Nucl. Med. Commun. 15 (2), 114–118. Kletter, K., et al., 1999. Radiation burden for the staff from pet and conventional nuclear medicine. Eur. J. Nucl. Med. 26 (9), OS358. Lucena, E.A., et al., 2007. Evaluation of internal exposure of nuclear medicine staff through in vivo and in vitro bioassay techniques. Radiat. Prot. Dosim. 127, 465–468. Mcfarland, R.C., 1991. Geometric consideration in the calibration of germanium detectors for filter-paper counting. Radioact. Radiochem. 2, 1. Ramos, V.S., et al., 2013. New filter for iodine applied in nuclear medicine services. Appl. Radiat. Isot.0 82, 111–118. Sales, F.G., 1981, Investigações sobre a remoção de compostos de iodo penetrantes na forma gasosa em materiais absorvedores. Tese de Mestrado, Departamento de Energia Nuclear – UFPE. Schwartz, I.G., et al., 1991. The risk of radiation exposure to laboratory personnel. Lab Med. 22 (2), 114–119. Vidal, M.S., et al., 2007. A methodology for auto-monitoring of internal contamination by I-131 in nuclear medicine workers. Radiat. Prot. Dosim. 125 (1– 4), 483–487.
Contents lists available at ScienceDirect
Applied Radiation and Isotopes journal homepage: www.elsevier.com/locate/apradiso
Internal dosimetry of nuclear medicine workers through the analysis of 131I in aerosols Luana Gomes Carneiro n, Eder Augusto de Lucena, Camilla da Silva Sampaio, Ana Letícia Almeida Dantas, Wanderson Oliveira Sousa, Maristela Souza Santos, Bernardo Maranhão Dantas Instituto de Radioproteção e Dosimetria (IRD-CNEN), Av. Salvador Allende s/n, 22783-127 Rio de Janeiro, Brasil
H I G H L I G H T S
The analysis of aerosols in the workplace is recognized as a useful means for controlling and optimizing workers exposure involved in the handling of 131I. Volatilization of 131I occurs during handling of the solutions. Improved ventilation conditions would certainly reduce risks of incorporation.
art ic l e i nf o
a b s t r a c t
Article history: Received 31 July 2014 Received in revised form 12 November 2014 Accepted 24 November 2014 Available online 3 December 2014
131
Keywords: Iodine-131 Aerosols Nuclear medicine
I is widely used in nuclear medicine for diagnostic and therapy of thyroid diseases. Depending of workplace safety conditions, routine handling of this radionuclide may result in a significant risk of exposure of the workers subject to chronic intake by inhalation of aerosols. A previous study including in vivo and in vitro measurements performed recently among nuclear medicine personnel in Brazil showed the occurrence of 131I incorporation by workers involved in the handling of solutions used for radioiodine therapy. The present work describes the development, optimization and application of a methodology to collect and analyze aerosol samples aiming to assess internal doses based on the activity of 131I present in a radiopharmacy laboratory. Portable samplers were positioned at one meter distant from the place where non-sealed liquid sources of 131I are handled. Samples were collected over 1 h using high-efficiency filters containing activated carbon and analyzed by gamma spectrometry with a high-purity germanium detection system. Results have shown that, although a fume hood is available in the laboratory, 131I in the form of vapor was detected in the workplace. The average activity concentration was found to be of 7.4 Bq/m3. This value is about three orders of magnitude below the Derived Air Concentration (DAC) of 8.4 kBq/m3. Assuming that the worker is exposed by inhalation of iodine vapor during 1 h, 131I concentration detected corresponds to an intake of 3.6 Bq which results in a committed effective dose of 7.13 10 5 mSv. These results show that the radiopharmacy laboratory evaluated is safe in terms of internal exposure of the workers. However it is recommended that the presence of 131I should be periodically re-assessed since it may increase individual effective doses. It should also be pointed out that the results obtained so far reflect a survey carried out in a specific workplace. Thus, it is suggested to apply the methodology developed in this work to other nuclear medicine services where significant activities of 131I are routinely handled as an effective means to optimize individual exposures and improve occupational radiation protection safety. & 2014 Elsevier Ltd. All rights reserved.
1. Introduction 131
I is one of the most used radionuclides in Nuclear Medicine due to its high affinity to the thyroid gland tissue, energy and type of n
Corresponding author. E-mail address: [email protected] (L.G. Carneiro).
http://dx.doi.org/10.1016/j.apradiso.2014.11.021 0969-8043/& 2014 Elsevier Ltd. All rights reserved.
radiation emitted. It is used in several procedures including diagnosis and treatment of diseases associated with thyroid. The handling of solutions containing radioiodine can result in a significant risk of occupational exposure due to the possibility of incorporation via inhalation or ingestion (IAEA, 1999; Schwartz et al., 1991). Several studies have been published about occupational exposure in nuclear medicine procedures, where increased use of
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
radioisotopes and availability of suitable methodologies for individual dose assessment have been discussed (Baechler et al., 2011; Bento et al., 2012; Dantas et al., 2007; Dumeau et al., 2011; Heller, 1996; Jansen et al., 1994; Kletter et al., 1999; Vidal et al., 2007). A previous study including in vivo and in vitro methods performed by Lucena et al. (2007), showed the occurrence of 131I incorporation by workers involved in handling of solutions used for radioiodine therapy. The International Atomic Energy Agency recommends, in its Safety Guide Series No RS-G-1.2 (IAEA, 1999), to implement occupational monitoring where the risk of annual effective dose resulting from intakes of radionuclides exceeds the value of 1 mSv. Such document suggests evaluating the workplace through the analysis of aerosols since inhalation is one of the main routes of entry of radionuclides in the human body and the main route of occupational exposure by 131I. This indirect measurement method have a considerable advantage compared to others, since the worker do not have to move from the workplace. Furthermore, it is possible to analyze a great number of samples simultaneously. Due to Brazilian continental dimensions, direct methods of monitoring (in vivo bioassay) is unfeasible if all measurements have to be performed at the institutes of the National Commission of Nuclear Energy (CNEN). In this case aerosol analysis should be the most suitable method for evaluation of these workplaces. In general, elements found in vapor form are not collected efficiently with conventional filters since porosity is not sufficient for its retention. In the case of iodine, specific sorbent materials are used to retain the element. Several research groups have tested methods to collect airborne 131I and activated charcoal was found to be one of the most efficient materials to adsorb iodine because the process is favored by the intrinsic nature of the substrate, i.e., its high porosity and high surface resulting area of the charcoal. Ramos et al. (2013) tested the efficiency of activated carbon for the adsorption of I2, the most common form of volatile iodine found in nuclear medicine, obtaining an efficiency of 100% retention of this chemical form. According to Sales (1981), the use of activated charcoal as a filter medium should be controlled due to its low flash point, so, the smaller the amount of charcoal the greater safety. The analysis of aerosols in the workplace is recognized as a useful means for controlling and optimizing workers exposure involved in the handling of 131I since direct methods are not usually available in all internal dosimetry laboratories. Thus, the aim of this work is to develop a method using activated carbon filters to assess the presence of 131I in aerosol samples collected in Nuclear Medicine Services as a subside to assess internal doses associated to the incorporation of this radionuclide by workers.
2.1. Preparation of filters and experimental arrangement Radiopharmaceutical laboratories usually have reduced dimensions and thus handling of radioactive materials requires high level of attention and care by the professional. Therefore, experimental arrangement was planned in order to cause the least possible inconvenience to routine work. The collection was carried out using a SKC portable Sampler Airchek model 224-44XR and a 37 mm diameter holder with filters containing activated carbon prepared in the Laboratory of Aerosol Characterization of the IRD. The filters, so-called “activated charcoal sandwich”, were prepared using the following materials: 37 mm diameter glass fiber filters; activated charcoal with a granulometry o100 μ and white glue. For each filter, 0.15 mg of charcoal was weighed on a glass fiber filter, and sealed in a sandwich form using white glue applied on the upper edge of the filter. After drying, the ready-made filters are placed in a holder coupled in the suction pump at low flow. Fig. 1 shows the Samplers Airchek and holder. A schematic representation of the samplers positioning is shown in Figs. 2 and 3. It consisted in four samplers, two at the right and two at the left side of the fume hood where 131I solution is handled. 2.2. Determination of
131
I in activated carbon filters
The filter samples containing 131I were analyzed at the In Vitro Bioassays Laboratory of IRD by gamma spectrometry with a Canberras HPGe coaxial detection system operated with the software Genies-2000. The detection system was calibrated with a 133Ba standard source certified by the Ionizing Radiation Metrology National Laboratory of IRD.
2. Materials and methods This study was performed in the Nuclear Medicine Service of a public Hospital located in the city of Rio de Janeiro, whose workers agreed to participate voluntarily in this research. A variety of radiopharmaceuticals are handled in this service, including 131I, the main interest of this work, in the form of capsules and liquid solutions. Since the manipulation of capsules is believed to represent low risk of internal exposure, the investigation was restricted to the radiopharmacy laboratory where iodine solutions are fractionated and doses are prepared to be administered to the patients. A total of sixteen samples were collected during four sampling procedures. The samples were analyzed by gamma spectrometry with Canberras HPGe coaxial semiconductor detector available at the In Vitro Bioassay Laboratory of the Dosimetry Division of the Institute of Radiation Protection and Dosimetry (IRD).
71
Fig. 1. SKC Sampler Airchek model 224-44XR and holder.
72
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
Fig. 2. Schematic representation of the samplers positioning related to the fume hood where
A counting geometry so-called “sandwich” was established in which 0.15 g of an activated carbon was spiked with the 133Ba standard source by gravimetric transference from a plastic vial and sealed between two glass fiber filters (Mcfarland, 1991). The solution was added to the substrate following a geometric model based on hexagonal geometric scheme proposed by Ceccatelli et al. (2010). The activity present in the source at the time of the measurements was corrected according to Eq. (1): − λt
A = A0 x e
(1)
The 131I equivalent activity of the source was calculated according to Eq. (2):
131
I=
131
I solutions are handled by workers.
A131Ba × ∑ γ 133Ba ∑ γ 131I
(2)
The calibration factor was obtained according to Eq. (3):
Fc =
cpm (filter) AEq
(3)
where cpm is the net count rate (count rate in the region of interest subtracted from background count rate), AEq is the 131I equivalent activity of the filter. The calibration of the detection system allows determining 131I activities and concentrations in filter samples.
Fig. 3. Sampling setup: Air filter and pump positioning are planned in order to cause minimum interference on routine work.
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
Table 1 Results for airborne samplings in a radiopharmacy laboratory of a nuclear medicine service in Rio de Janeiro. Sample
1 2 3 4 5 6 7 8 9 10 11 12 13 14 15 16 Average
Collection point
Right (outer face fume hood) Left (outer face fume hood) Right (1 m) Left (1 m) Right (1 m) Right (outer face fume hood) Left (outer face fume hood) Left (1 m) Right (1 m) Right (outer face fume hood) Left (outer face fume hood) Left (1 m) Right (1 m) Right (outer face fume hood) Left (outer face fume hood) Left (1 m) concentration
Filter activity (Bq)
Concentration (Bq/m3)
7.5E þ00 1.7E þ00 2.3E þ00 2.3E þ00 4.5E 03 2.5E 02 7.5E 03 7.4E 03 1.33E 02 3.22E 03 1.20E 02 3.91E 03 o MDA o MDA o MDA o MDA 3 Bq/m3
1.9E þ 01 4.4E þ00 5.8E þ00 5.9E þ00 3.0E 02 1.4E 01 4.5E 02 5.3E 02 3.4E 02 8.2E 03 2.6E 02 7.7E 03 o MDA o MDA o MDA o MDA
n
Minimum Detectable Activity (MDA) ¼ 1.60E 02 Bq.
73
defined as that concentration of airborne activity which would result in the intake of Ij,inh,L by a worker exposed continuously for one year (taken to be 2000 working hours), being Iinhl,L an intake corresponding to the relevant limit on effective dose, e.g., 20 mSv (IAEA, 1999). For a standard breathing rate of 1.2 m3/h, the DAC would thus be given by Eq. (4):
DAC =
I j, inh, L (4)
2000 × 1.2
Applying specific parameters related to 131I in vapor form, DAC is calculated as 416.67 Bq/m³. The experimental data obtained in this work indicate that the exposure estimated over one year of work, for different concentrations are all below the values of DAC. However, it should be taken into consideration that these workers are involved in the handling of other radiopharmaceuticals in the same service, and that this exposure is chronic. It is therefore justifiable to implement actions aiming to reduce 131I incorporations and consequently individual occupational exposures and radiological risks.
3. Results The collection of airborne samples performed at the Radiopharmacy Laboratory investigated in this work used activated carbon filters and an operation flow of 1 L/min. The results obtained are presented in Table 1. The DAC value for 131I in vapor form of 416.67 Bq/m³ is calculated considering a routine workload of 8 h per day and 5 days per week, which results in 2000 h per year of constant exposure over the whole period. However the workers involved in the handling of radiopharmaceuticals do not comply with such regime. Actually the staff perform the tasks in an alternate basis in order to minimize handling and consequently to reduce individual exposure. Based on this actual differentiated work routine, DAC values for different exposure times were calculated considering the manipulation taking place twice a week, and a handling time registered by the workers according to their real work routine. Table 2 presents the parameters used to estimate concentrations under such conditions, and considering the mean activity values for DAC.
4. Discussion The concentration of radionuclides in air is evaluated by the Derived Air Concentration (DAC, expressed in Bq/m3) which is
Table 2 Derived air concentrations of
5. Conclusions The procedures developed in this work and applied for sample collection and radiometric analysis were suitable to the objective of the investigation, both in terms of application easiness and sensitivity required for an assessment of occupational exposure. Based on the results obtained it can be concluded that volatilization of 131I occurs during handling of the solutions. Incorporation was experimentally confirmed through additional monitoring of thyroid and urine of workers involved in the handling, improving sensitivity and applicability of the method to this group of workers. The calculation of the DAC for different working conditions suggests the efficacy of reducing exposure time. Furthermore, improved ventilation conditions would certainly reduce risks of incorporation. Additional studies are needed to verify whether the presence of 131 I in the air is associated with good practices, laboratory design or solely physical and chemical properties of the radionuclide itself. Such information would guide specific recommendations on the improvement of radiological protection conditions of the workplace.
131
I as a function of the time of exposure.
Exposure time (min)
Exposure time (h)
Exposure time/week (h)
Exposure time/year (h)
DAC (Bq/m3)
Concentration (Bq/m3)
10 20 30 40 50 60 70 80 90 100 110 120
0.17 0.33 0.50 0.67 0.83 1.00 1.17 1.33 1.50 1.67 1.83 2.00
0.33 0.67 1.00 1.33 1.67 2.00 2.33 2.67 3.00 3.33 3.67 4.00
17.33 34.67 52.00 69.33 86.67 104.00 121.33 138.67 156.00 173.33 190.67 208.00
4.86E þ04 2.43E þ04 1.62E þ 04 1.21E þ 04 9.71E þ 03 8.09E þ03 6.94E þ03 6.07Eþ03 5.40E þ03 4.86E þ03 4.41E þ03 4.05E þ03
8.67Eþ 00 1.73E þ 01 2.60E þ 01 3.47E þ 01 4.33Eþ 01 5.20E þ 01 6.07Eþ 01 6.93E þ 01 7.80E þ01 8.67Eþ 01 9.53E þ 01 1.04E þ02
74
L.G. Carneiro et al. / Applied Radiation and Isotopes 100 (2015) 70–74
Acknowledgments The authors acknowledge the CNPq for the financial support (Grant number 503563/2012-9). We especially thank the Director of the Hospital, the Head of the Division of Nuclear Medicine and the professionals who participated voluntarily in this research.
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