Introduction: Active Personal Dosimeters (APDs) are well accepted as useful and reliable instruments for individual dosimetry measurements. APDs have many advantages compared with passive Dosimeters for individual external radiation dose assessments. In routine monitoring, occupational exposure is carried out for verifi-cation and demonstration of compliance with the regulatory dose limits. So, it is one of the most important tools in order to achieve or demonstrate the level of radiation protection. Methods: Yazd province has only one private nuclear medicine (NM) center. In this center, two NM technologists exposed to radioactive patients during radiophar-macuticals preparation were monitored. NM technologists have to be close to the patient during radiopharmaceutical injection and patient positioning on the gamma camera table. An electronic Personal Dosimeter DKG-21 Ecotest made in Ukraine which records the ambient dose equivalent rate and equivalent dose was used to monitor the radiation exposure to the technologists and to record the accumulation dose in mSv throughout a working day. This study was accomplished between the time period of January to June 2011. The Dosimeter is designed to measure indi-vidual equivalent dose Hp(10). The dose range of gamma radiation was 0.01 mSv to 1 Sv and the energy range 0.05 to 6 MeV which was suitable for NM procedures. The planar and tomography NM images were performed by the 2 technologists in the morning and afternoon shifts. Results: The average monthly occupational dose of each technologist was ap-proximately 0.6 mSv. Their annual doses were 6.6 and 8.8 mSv, respectively. They were lower than the maximum permissible dose of 20 mSv/y. Total number of NM procedures performed in this NM center during June 2010 to June 2011 was 3265. Conclusion: The use of APD for monitoring the NM technologists is a use¬ful tool to check compliance with regulatory dose limits and radiation protection principals.

In the present work performance of film badge as an alternative Personal Dosimeter for thermal neutrons is investigated. To do this, a cadmium-lead (Cd-Pb) filter with the same thickness as the tin-lead (Sn-Pb) filter is attached to the AERE/RPS badge. Since thermal neutrons are mixed with gamma rays, the Dosimeter is irradiated by the Co-60 gamma rays standard field of Karaj Secondary Standard Dosimetry Lab as well as the mixed neutron-gamma field of the radiography beamline of Isfahan Miniature Neutron Source Reactor. In the both fields, ten Personal dose-equivalent values between 0. 1 to 10 mSv are chosen. For any dose, three film badges are used and the net optical density is determined as the average of their optical densities. Finally, the calibration curves of the film badge are plotted to determine the dose-equivalent values. Obtained results reveal that film badge simultaneously determines the thermal neutrons and gamma rays dose fractions. Also, the thermal neutron doses are at most 50% different from the nominal values considered.

Background: Neutron contamination in our environment can cause harmful bio-logical effects on human body. Therefore, many efforts have been made to construct a neutron Dosimeter to estimate the received dose. Objective: To design a simple neutron Dosimeter. Methods: The primary Dosimeter had 3He as a spherical thermal neutron detector encircled by paraffin 10 cm in radius. Then, the paraffin sphere was replaced with ICRU that contains soft tissue and dose equivalent determined as a desired output. Finally, an appropriate relation between counts and dose equivalent was found. Results: Results on the energies below 1 MeV demonstrated the similarity of changing process of these two quantities, so they could relate to each other with an adequate factor. To find the best fit, different factors considered and the smallest c2 (goodness of fit) was 1.17×105 At the next step, two covers of cadmium and gado-linium, separately, put around the detector to improve c2, which was 2.51×104 for cadmium cover and 6.33×103 for gadolinium cover. As we see, gadolinium cover fits the curves of counts and equivalent dose in a better way. Conclusion: Applying this simple Dosimeter lead us to estimate whole body dose equivalent.

Background: Thermoluminescence dosimetry is one of the dosimetry procedures used widely as routine and Personal Dosimeters. In order to extend this kind of Dosimeters, dental tissue has been examined and was found promising as a TLD Dosimeter. Materials and Methods : In this study, 70 health teeth were collected. The only criterion, which was considered for selection of the teeth, was the healthiness of them regardless of age and gender of the donors. All collected samples were washed and cleaned and milled uniformly. The final powder had a uniform grain size between 100 – 300 micrometer. The sample was divided into four groups. Group A and B were used for measurement of density and investigation of variation of thermoluminescent characteristics with temperature respectively. Groups C and D were used for investigation of variation of thermoluminescent intensity with dose and fading of this intensity with time . In all cases the results obtained with dental tissue were compared to a standard LiF, TLD Dosimeter. Results : It was found that, average density of the dental tissue was 1.570 g/cm3, which is comparable to density of LiF, which is 1.612 g/cm3. It was also concluded that the range of 0-300 oC, dental tissue has a simple curve with two specific peaks at 140 and 250 oC respectively.The experiment also showed that, the variation of relative intensity versus dose is linear in the range of 0.04 – 0.1Gy . The fading rate of dental tissue is higher than LiF but still in the acceptable range (14% per month in compare to 5.2% per month) Conclusion : Dental tissue as a natural Dosimeter is comparable with TLD and can be used in accidental events with a good approximation