Calibration of gamma cameras for the evaluation of accidental intakes of high-energy photon emitting radionuclides by humans based on urine samples

The prompt response to emergency situations involving suspicion of intakes of radionuclides requires the use of simple and rapid methods of internal monitoring of the exposed individuals. The use of gamma cameras to estimate intakes and committed doses was investigated by the Centers for Disease Control and Preventions (CDC) of the USA in 2010.The present study aims to develop a calibration protocol for gamma cameras to be applied on internal monitoring based on urine samples to evaluate the incorporation of high-energy photon emitting radionuclides in emergency situations. A gamma camera available in a public hospital located in the city of Rio de Janeiro was calibrated using a standard liquid source of Eu supplied by the LNMRI of the IRD. “Efficiency vs Energy” curves at 10 and 30 cm were obtained. Calibration factors, Minimum Detectable Activities and Minimum Detectable Effective Doses of the gamma camera were calculated for Cs and Co. The gamma camera evaluated in this work presents enough sensitivity to detect activities of such radionuclides at dose levels suitable to assess suspected accidental intakes.


INTRODUCTION
In Brazil, as in many countries, there is a significant number of radioactive sources and, in many situations access control is still ineffective, giving rise to accidents involving the loss and theft of sources.The Brazilian Nuclear Regulatory Board (CNEN) is responsible for authorizing and auditing industrial and medical facilities as well as research and educational institutions where a variety of open sources of radionuclides are routinely handled [1].
The feasibility of the use of gamma cameras for the evaluation of the incorporation of radionuclides by nuclear medicine workers was studied by Dantas et al. [2] in the scope of the IAEA-TC-RLA 049 Project.The US-CDC (Centers for Disease Control and Prevention) also published a document containing instructions for the use of gamma cameras for calculating radioisotope activity in the human body [3].The calibration of this type of equipment requires the determination of the detector efficiency and the estimation of the minimum detectable activity for each radionuclide of interest.
A preliminary evaluation of the severity of the accident may be carried out in nuclear medicine services participating in a network to be established for this purpose.Thus, the aim of this study is to provide a protocol for the evaluation of intakes of radionuclides by humans in emergency situations through the application of an in vitro bioassay method.

MATERIALS AND METHODS
The evaluation of internal exposure in practices with risk of intakes of radionuclides requires the application of in vivo and in vitro monitoring techniques, as well as methodologies for the interpretation of bioassay data [4].Measurements are usually performed with scintillation or semiconductor detectors depending on radionuclide emissions, and require the use of radioactive standards for the calibration and quality control of the detection systems.
The evaluation of the sensitivity of the detection system is based on its minimum detectable activities for the radionuclides of interest in comparison to the expected activity in the organs and tissues to be measured.Such evaluation relies on retention and excretions factors in the body compartments as a function of the time elapsed between intake and measurement.The AIDE Software [5] performs the calculations necessary to provide activities in organs of interest as well as the committed effective dose for a wide variety of radionuclides in different intake patterns, chemical and physical forms.
This work was performed in a public military hospital where a gamma camera Phillips BrightView-XCT was calibrated with a standard liquid source of 152 Eu provided by the National Laboratory for Metrology of Ionizing Radiation (LNMRI-IRD).The source was uniformly distributed in an acid solution contained in a 1L volume polyethylene bottle.
Five consecutive measurements were performed at 2 and 10 cm distance from the source to the front face of the detector, and the count rates were recorded in the regions of interest corresponding to the photons of 40.1, 344.3 and 778.9 keV of 152 Eu.The detection efficiency in each ROI was calculated as follows: where Eff is the detection efficiency in the ROI of the photon of interest; C is the total counts in the ROI; T is the count time; A is the activity of the standard source of 152 Eu and Ig Ig is the photon yield at the measured energy.
The software used to control the gamma cameras acquisition allows to record counts in the ROIs using the "spectrum mode".
As shown in Figure 1, a spacer made of Styrofoam was used to allow correct positioning of the plastic bottle in relation to the front face of the gamma camera.It should be highlighted that, in order to increase the sensitivity, all measurements where performed after removing the high energy collimator of the gamma camera.
The Efficiency x Energy curves obtained at 2 and 10 cm were used to calculate the efficiencies of the gamma camera to specific radionuclides of interest.In this work efficiencies were calculated for the 364 and 662 keV of 131 I and 137 Cs respectively.Subsequently the Minimum Detectable Activities (MDA) were calculated for the same radionuclides, which are assumed to be of concern in situations related to radiological and nuclear accidents.
Minimum Detectable Activities for 131 I and 137 Cs were calculated at each geometry based on respective room background of the gamma camera at the ROIs and efficiency values as follows: where MDA is the Minimum Detectable Activity; N is the total counts of the background in the ROI in 1 minute; Eff is the detection efficiency (cps/dps) and Ig is the photon yield at the measured energy.( where MDA is the minimum detectable activity and m(t) is the excretion fraction urine.
Using the MDI value and the dose coefficient given by AIDE software, it was calculated the Minimum Detectable Effective Dose to both radionuclides.
(4)   Table 1 presents Efficiency values of the gamma camera for the measurement of 131 I and 137 Cs in urine samples positions at 2 and 10 cm distances from source to detector, calculated based on the Efficiency x Energy curves obtained with the standard source of 152 Eu.The minimal detectable activities for 131 I and 137 Cs are also shown in Table 1.The gamma camera proved to be a very sensitive device in terms of measurable activity since the MDA values in 1 minute count time are all below 100 Bq both for 131 I and 137 Cs at 2 and 10 cm distances adopted in this study.

RESULTS AND DISCUSSION
Table 2 shows the corresponding MDIs and MDEDs values calculated as a function of the excretion rates provided by the AIDE software in the simulated intake scenario and the MDAs calculated based on the calibration results.The efficiency vs energy curve obtained with 152 Eu was used to calculate detection efficiencies of the gamma camera for 131 I and 137 Cs for evaluation purposes in this work.However it is also important to point out that it is possible to use the same curve to determine the activity content in urine samples provided by individuals potentially exposed to any photon emitting radionuclide in the energy range covered by this calibration curve.

Figure 1 :
Figure 1: Calibration geometry: Plastic Bottle containing standard source of 1 52 Eu positioned at 10 cm distance to the gamma camera front face

Figures 2 and 3
Figures 2 and 3 show the efficiency vs energy curves obtained using three ROIs of 152 Eu at 2 and 10 cm distance source-detector respectively.

Figure 2 :
Figure 2:-Efficiency curves of 152 Eu at 2 cm distance from source to detector.

Figure 3 :
Figure 3: Efficiency curve of 152 Eu at 10 cm distance from source to detector.

Table 1 :
Efficiencies and Minimum Detectable Activities of the Gamma Camera for the measurement of 131 I and 137 Cs in 1 L urine samples at 2 and 10 cm distances

Table 2 :
Minimum Detectable Intakes (MDI) and Minimum Detectable Effective Doses (MDED) of the Gamma Camera for the measurement of 131 I and 137 Cs in 1 L urine samples at 2 and 10 cm distances Furthermore, according to the IAEA, in certain circumstances, in emergency situations it is tolerated that individual doses could reach higher values.Therefore, this equipment can be considered suitable for monitoring radionuclide intakes in such situations.