Determination of uncertainties associated to the in vivo measurement of iodine-131 in the thyroid.

Intakes of radionuclides can be estimated through in vivo measurements, and the uncertainties associated to the measured activities should be clearly stated in monitoring program reports. This study aims to evaluate the uncertainties of in vivo monitoring of iodine 131 in the thyroid. The reference values for high-energy photons are based on the IDEAS Guide. Measurements were performed at the In Vivo Monitoring Laboratory of the Institute of Radiation Protection and Dosimetry (IRD) and at the Internal Dosimetry Laboratory of the Regional Center of Nuclear Sciences (CRCN-NE). In both institutions, the experiment was performed using a NaI(Tl) 3''3″ scintillation detector and a neck-thyroid phantom. Scattering factors were calculated and compared in different counting geometries. The results show that the technique produces reproducibility equivalent to the values suggested in the IDEAS Guide and measurement uncertainties is comparable to international quality standards for this type of in vivo monitoring.

A mobile bioassay laboratory for the assessment of internal doses based on in vivo and in vitro measurements.

Internal exposures may occur in nuclear power plants, radioisotope production, and in medicine and research laboratories. Such practices require quick response in case of accidents of a wide range of magnitudes. This work presents the design and calibration of a mobile laboratory for the assessment of accidents involving workers and the population as well as for routine monitoring. The system was set up in a truck with internal dimensions of 3.30 m × 1.60 m × 1.70 m and can identify photon emitters in the energy range of 100-3,000 keV in the whole body, organs, and in urine. A thyroid monitor consisting of a lead-collimated NaI(Tl)3" × 3" (7.62 × 7.62 cm) detector was calibrated with a neck-thyroid phantom developed at the IRD (Instituto de Radioproteção e Dosimetria). Whole body measurements were performed with a NaI(Tl)8" × 4" (20.32 × 10.16 cm) detector calibrated with a plastic-bottle phantom. Urine samples were measured with another NaI(Tl) 3" × 3" (7.62 × 7.62 cm) detector set up in a steel support. Standard solutions were provided by the National Laboratory for Metrology of Ionizing Radiation of the IRD. Urine measurements are based on a calibration of efficiency vs. energy for standard volumes. Detection limits were converted to minimum committed effective doses for the radionuclides of interest using standard biokinetic and dosimetric models in order to evaluate the applicability and limitations of the system. Sensitivities for high-energy activation and fission products show that the system is suitable for use in emergency and routine monitoring of individuals under risk of internal exposure by such radionuclides.

Estimation of internal exposure to 99Mo in nuclear medicine patients.

(99m)Tc is the most widely used radionuclide in nuclear medicine. It is obtained by elution of (99)Mo-(99m)Tc generators. Depending on the quality of the generator and its integrity, (99)Mo may be extracted from the column during the elution process, becoming a radionuclidic impurity in the (99m)Tc eluate. This fact would impart an unnecessary dose to the patients submitted to diagnostic procedures. The aim of this work is to evaluate (99)Mo incorporation and internal effective doses in nuclear medicine patients through bioassay techniques, providing information on the metabolism of molybdenum in humans. A methodology based on in vivo and in vitro measurements was developed. In vivo measurements were performed with a NaI detector installed in the IRD WBC. Urine samples were analysed with a HPGe at the IRD bioassay laboratory. Patients showed detectable activities of (99)Mo in whole body and urine. Results were interpreted with AIDE software. Estimated incorporation was compared to predicted values based on ICRP model. Effective doses were in the order of micro sieverts. Results suggest the need to implement a routine quality control program of radionuclidic impurity of (99)Mo in (99m)Tc eluates to be conducted by radiopharmacy laboratories of nuclear medicine centers.

Optimization of (131)I doses for the treatment of hyperthyroidism.

Several methods can be used to determine the activity of (131)I in the treatment of hyperthyroidism. However, many of them do not consider all the parameters necessary for optimum dose calculation. The relationship between the dose absorbed by the thyroid and the activity administered depends basically on three parameters: organ mass, iodine uptake and effective half-life of iodine in the thyroid. Such parameters should be individually determined for each patient in order to optimize the administered activity. The objective of this work is to develop a methodology for individualized treatment with (131)I in patients with hyperthyroidism of the Grave's Disease. A neck-thyroid phantom developed at the IRD was used to calibrate a scintillation camera and a uptake probe SCT-13004 at the Nuclear Medicine Center of the University Hospital of Rio de Janeiro and a uptake probe SCT-13002, available at the Nuclear Medicine Institute in Goiânia. The biokinetic parameters were determined based on measurements performed in eight voluntary patients. It is concluded that the use of the equipment available at the hospital (scintillation camera and uptake probe) has shown to be a suitable and feasible procedure for dose optimization in terms of effectiveness, simplicity and cost.

Evaluation of internal exposure of nuclear medicine staff through in vivo and in vitro bioassay techniques.

The manipulation of a wide variety of unsealed sources in Nuclear Medicine results in a significant risk of internal exposure of the workers. 131I should be highlighted among the most frequently used radionuclides because of its large application for diagnosis and therapy of thyroid diseases. The increasing use of radionuclides for medical purposes creates a demand for feasible methodologies to perform occupational control of internal contamination. Currently in Brazil, there are approximately 300 nuclear medicine centres in operation but individual monitoring is still restricted to the control of external exposure. This work presents the development of in vivo and in vitro bioassay techniques aimed to quantify incorporation of radionuclides used in Nuclear Medicine. It is also presented the results of a preliminary survey of internal exposure of a group of workers involved in the preparation of therapeutic doses of 131I. Workers were monitored with a gamma camera available in the Nuclear Medicine Service of the University Hospital of Rio de Janeiro and at the Institute of Radiation Protection and Dosimetry Whole-Body Counter (IRD-WBC). The in vivo detection systems were calibrated with a neck-thyroid phantom developed in IRD. Urine samples from radiopharmacy workers were collected after preparation and administration of therapeutic doses (10-250 mCi) of 131I and measured with a HPGe detection system available in the Bioassay Laboratory of IRD. The results show that the bioassay methods developed in this work present enough sensitivity for routine monitoring of nuclear medicine workers. All workers monitored in this survey presented positive results for 131I in urine samples and two workers presented detectable activities in thyroid when measured at the IRD-WBC. The highest committed effective dose per preparation was estimated to be 17 microSv.

A protocol for the calibration of gamma cameras to estimate internal contamination in emergency situations.

The concern about accidents involving radioactive materials has led to the search of alternative methods to quickly identify and quantify radionuclides in workers and in the population. One of the options to face up an eventual demand for mass monitoring of internal contamination is the use of a nuclear medicine diagnostic equipment known as gamma camera, a device used to scan patients who have been administered specific amounts of radioactive materials for medical purposes. Although the gamma camera is used for image diagnosis, it can be calibrated with anthropomorphic phantoms or point sources for the quantification of radionuclide activities in the human body. This work presents a protocol for the calibration of gamma cameras for such application. In order to evaluate the suitability of this type of equipment, a gamma camera available in a public hospital located in Rio de Janeiro was calibrated for the in vivo measurement of 131I. The calibration includes the determination of detection efficiencies and minimum detectable activities for each radionuclide. The results show that the gamma camera presents enough sensitivity to detect activity levels corresponding to effective doses below 1 mSv. The protocol is the basis to establish a network of Nuclear Medicine Centres, located in public hospitals in eight countries of Latin America (Argentina, Brazil, Colombia, Cuba, Chile, Mexico, Peru and Uruguay) and in Spain that could be requested to collaborate in remediation actions in the event of an accident involving incorporation of radioactive materials. This protocol is one of the most significant outputs of the IAEA-ARCAL Project (RLA/9/049-LXXVIII) aimed to the Harmonization of Internal Dosimetry Procedures.

Design and operation of a whole-body monitoring system for the Goiânia radiation accident.

With as many individuals involved in the Goiânia 137Cs accident who had high levels of internal contamination, it was necessary to improvise a whole-body counter installation in loco. The in-vivo counting system was located in a 4.0 X 3.5 X 3.5-m room, where seven layers of 2-mm lead sheets with dimensions of 2.0 m X 1.0 m were overlaid on the floor at loci that were equidistant from the walls. A 20-cm diameter NaI (Tl) detector was installed at a height of 2.05 m above the floor at the center of the room. The detector was shielded and collimated with 5 cm of lead. The enormous amounts of activity in the subjects required the detector to be positioned at a height of 2.05 m. Subjects were required to wear disposable clothing and lie on a reclining, fiberglass chair. Counting time for the subjects was 2 min (live-time). The minimum detectable 137Cs activity for this counting time was 7.3 kBq* (0.05 significance level). Besides the accident victims, all individuals who had direct or indirect contact with contaminated people or areas were also monitored. More than 300 people of both sexes, with ages varying from a few months to 72 y, were measured for whole-body radioactivity. The observed activities ranged from less than the minimum detectable activity (MDA) to 59 MBq.