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Year : 2013  |  Volume : 28  |  Issue : 5  |  Page : 21-24  

Radiation Safety

Date of Web Publication29-Nov-2013

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How to cite this article:
. Radiation Safety. Indian J Nucl Med 2013;28, Suppl S1:21-4

How to cite this URL:
. Radiation Safety. Indian J Nucl Med [serial online] 2013 [cited 2022 May 19];28, Suppl S1:21-4. Available from:


Estimation of radiation dose received by the radiation worker during nuclear medicine procedures: A preliminary study

Subhash Chand Kheruka, Vandana Kumar Dhingra 1 , Nisha Bhatia 1

Departments of Nuclear Medicine, SGPGIMS Lucknow, Uttar Pradesh, 1 Cancer Research Institute, HIHT, Dehradun, Uttarakhand, India

Objective: The radiation dosimetric literature concerning the medical and non-medical personnel working in nuclear medicine departments are limited and are of concern for radiation personnel. We aimed to measure the radiation dose received by the staff involved in a diagnostic (non-PET) nuclear medicine department having a single gamma camera (SPECT). Materials and Methods: The effective whole body doses to the radiation workers over a period of 2 years were evaluated based on personal monitoring dose report using thermoluminiscence dosimeter (CaSo4:Dy) of chest and wrist badges. Results: The department comprised of one physician, technologist and assistant during the study period with 3000 patient studies (Tc99m and low dose radioiodine). An average of 200 mCi of Tc99m fortnightly and 25 mCi of I-131 monthly were used. The total whole body doses received by the nuclear medicine physician, technologist and support staff over the study period were 0.5 mSv, 3.75 mSv and 2.2 mSv respectively. considering the uniform activity distribution over all the patient studies the whole body dose received by the personnel per patient study were 0.167 uSv, 1.25 uSv and 0.73 uSv respectively. The dose to the extremity as recorded by wrist badge of the nuclear medicine physician and technologist were 0.15 mSv and 26.8 mSv respectively with an average overall per patient study were 0.05 uSv and 1.25 uSv respectively. Conclusion: This study reveals that by following the basic radiation protection rules the personal doses received by staff of a typical diagnostic (non-PET) nuclear medicine department is well within the permissible limits, with maximum dose to the technologist and a rare possibility to receive doses exceeding the maximum permissible doses.


Improvement on lack of awareness and fear of radiation among general medical workers: Potential role of the nuclear medicine worker?

Vandana Kumar Dhingra, Subhash Chand Kheruka 1 , Nisha Bhatia

Departments of Nuclear Medicine, Cancer Research Institute, HIHT, Dehradun, Uttrakhand, 1 SGPGIMS, Lucknow,

Uttar Pradesh, India

Introduction: Today's medical practice involves all to commonly come across radiation facilities on a regular visit to the hospital. All medical workers would be dealing with these patients, thus they should be aware about radiation in medicine. Aim: In this preliminary study we evaluated awareness and fear of radiation amongst medical trainees and workers who are not working in radiation areas. Materials and Methods: A series of short workshops were conducted by nuclear medicine department in a medical college setting over 1 year. Participants included trainee doctors, nurses and technical workers not working in radiation areas. Results: Study included 200 participants in 8 sessions with 140 nurses, 40 doctors and 20 staff working in non-radiation areas. Most participants (N = 186 i.e > 90%) had incorrect idea about radiation. They were either too scared (N = 153) or too oblivious (N = 33) to radiation and its effects. Questionnaires revealed improvement in correct answers from <30% (before) to >70% (after). Workshop included information on basis of radiation, hazards radiation safety. Conclusion: We found a gross lack of awareness of any kind about medical radiation and the workshops made a significant difference in the level of awareness, confidence and understanding about basic radiation safety amongst medical workers. This has scope to be improved upon by radiation workers. We feel that if nuclear medicine workers take a leading role in this area this could additionally have a positive impact on perception of nuclear medicine procedures as most participants felt that nuclear medicine procedures were the most hazardous in general.


Estimation of dose to public from stack releases of medical cyclotron facilities

Shimja Bhanu, Manisha Inamdar, Pankaj Tandon, Bokka Rajshekharrao 1

Radiological Safety Division, Atomic Energy Regulatory Board, 1 Thyrocare Technologies Limited, Navi Mumbai, Maharashtra, India

Objective: The aim of this study is to estimate the dose received by the general public from stack releases of medical cyclotron facilities. Materials and Methods: This study covers all the functional medical cyclotron facilities in India where various positron emitting radiopharmaceuticals are routinely manufactured for positron emission tomography (PET) applications. The Guassian plume model is used to calculate the ground level air concentration, at downwind distance x, of the radio nuclides that can be present in the air due to environmental releases from these facilities. As F-18 is the major isotope produced in these facilities, the calculations are done for F-18 using the equation, where F is the Guassian distribution factor and Q is the release rate of F-18 from the stack. A value of 0.25 for (the fraction of the time during the year that the wind blows towards the receptor of interest) and 2 m/s for (the wind speed) is considered. In order to get a conservative estimate of the ground level air concentration, the calculations are repeated for the values 0.5 and 1 m/s for and respectively. Two different scenarios are considered for these calculations. 1 i.e. the stack height is greater than 2.5 times the height of the building that affect airflow near the release point and 2) and where represents the surface area of the largest wall of the building most influencing the plume flow. For the second scenario is calculated by, where is a constant of value 1 m. The values of are used to calculate the dose received by general public assuming that they are exposed to this pathway throughout the year. Results: The maximum value of the dose to general public obtained from the calculations based on different possible scenarios of medical cyclotron facilities is 0.028 μSv/year for a unit release rate (1 Bq/s) of activity from the stack. For a limiting dose value of 10 μSv/y, the maximum allowable release rate from the stack of medical cyclotron works out to be around 350 Bq/s. Conclusion: The value obtained in this approach is most conservative. In actual situations this value can be modified using the site specific parameters in the equations and also considering the fact that medical cyclotrons are not operated continuously.


Volatility of radioiodine from indigenously prepared high dose (150 mCi) Na 131 I capsule

GL Vispute, A Thulasidhasan, Anand Gaurav, Kiran S Mehra, Ravi Seshan, SS Sachdev

Radiopharmaceuticals Programme, Board of Radiation and Isotope Technology, Navi Mumbai, Maharashtra, India

Objective: To quantify the extent of volatilization of 131 I from a dispensed vial containing Na 131 I therapeutic capsule. While iodine as oral solution is beneficial to thyroid patients, the frequency of contamination of nuclear medicine personnel/workplace was noted frequently. This contamination and unnecessary exposure to occupational workers was overcome by formulating as capsules. Airborne contamination however still occurs with capsules and can be a source of concern. Encapsulation of 131 I capsules proven to be highly effective method of decreasing these risks. Materials and Methods: Therapeutic doses of Na 131 I for oral administration are made as "capsule in capsule" which involves encapsulating a smaller hard gelatin capsule (of size "1") containing 131 I activity dispensed on a solid support (anhydrous sodium sulphate), with a larger size (size "0") hard gelatin capsule. The completed capsule is transferred to glass vials. Charcoal pouch was placed on the mouth of the open vials to trap volatilized 131 I escaping from the 131 I therapy capsule. The radioactivity adsorbed by charcoal pouch was monitored for thirty days. The charcoal pouches were counted in a pre-calibrated dose calibrator. Conclusion: The amount of 131 I that volatilizes from Na 131 I therapy capsules is minimal. Only a small percentage of the activity present in the capsules is released as volatile. These results show that 131 I therapy capsules are safe and convenient for both the patient and nuclear medicine personnel. The capsule vial should be opened in a fume hood. This will allow the volatile airborne 131 I to get released in the gaseous waste and preventing the 131 I activity escaping into the room air.


Assessment of volatilization of radioiodine (I-131) during handling in the nuclear medicine hot lab

Om Prakash, R Tiwari, V Patel, S Rach, V Bhat, RK Vyas

Department of Nuclear Medicine, Gujrat Cancer and Research Institute, Ahmedabad, Gujarat, India

Introduction: Radioiodine was first introduced in the early 1940 as an oncologic therapeutic agent, till today it is agent of choice. I-131 is isotope of I-127 which is a beta and gamma emitter. Its maximum beta energy is 0.61 Mev and average energy is 0.192 Mev. In iodide form of radioiodine, the iodide ions are strong reducing agent, so it readily gives up one electron. Iodide solution may acquire a brownish tint as a result of oxidation of iodide to free iodine by atmospheric oxygen. This free iodine is volatile and comes in air as vapor. Objective: Our aim of study was to calculate the airborn activity due to radioiodine handling, is it safe to handle. Materials and Methods: Sodium iodide (NaI I-131) kept in a bicker which airtightly packed and one tube which contain charcoal filter was connected with foot operated suction instrument. Two same type of arrangement were made for liquid and capsule form of I-131. In one bicker 500 micro curie liquid and in other 500 micro curie capsule were placed. Two sequential charcoal filters sere changed after suction of air from bicker on day 1 st , 2 nd and 3 rd . The filters were placed in different tube and sealed it. All the tubes wee counted in well counter (captus 3000). Result: The filters of capsule shows <0.015% released activity of total amount. Gradually it decreases on 2 nd and 3 rd days. Liquid iodine filters counts were <0.04% of total activity which also gradually decrease on 2 nd and 3 rd days. Conclusion: I-131 capsule volatilization was minimal, only very small percentage of total amount released in air. The amount of I-131 released during patient dose administration is expected to be considerably very less, because the whole procedure take only few minutes. In our experiment we kept them 24 h open. So the results shows I-131 radiopharmaceutical compounds are safe and convenient for patients and worker. For following the ALARA principal we should handle them in fume hood.


Quantification of short-lived activity in air samples from medical cyclotron

Kamaldeep, Pravind Kumar 1 , Sushma G Kaisar 1 , Shriram Tervankar 1 , MGR Rajan 1 , RK Gopalakrishnan

Radiation Safety System Division, 1 Radiation Medicine Centre, BARC, Mumbai, Maharashtra, India

Objective: In medical cyclotron short lived isotopes (viz, 18 F, 13 N, 11 C, 15 O) are produced which are subsequently converted into radiopharmaceuticals by radiochemical procedures. During the radiochemical procedures there exists a potential of internal radiation dose to the occupational workers. Hence quantification of the activity in the air samples of these short-lived isotopes is important. Materials and Methods: Due to the short half lives of the radionuclides contributing to the air borne concentration in the cyclotron facility (half life of 18 F, 13 N, 11 C and 15 O are 110 min, 10 min, 20 min and 2 min respectively) the radionuclides decays significantly while it is being sampled. During sampling equilibrium between the amount collected and the amount decaying will be reached. The equilibrium amount of collected activity will be determined by the difference between the collection rate and the decay rate. For deriving an expression for estimation of activity for short-lived radionuclides, decays during sampling, delay in sample counting and during the counting of air sample were considered. Appropriate algorithms were developed for immediate assay of these short lived nuclides. Results: After considering the decays an expression was developed to quantify the activity present in air sample for short-lived radionuclides. The expression is function of sample counts, decay constant, efficiency of the detector, flow rate of the sampler and decay during sampling, delay time and counting. Conclusion: An easy to use algorithm is made to assess the airborne concentration in the medical cyclotron facility incorporating the decay corrections at all stages and is found to be very effective.


Radiometry of positron emission tomography-computed tomography facility, RMC

Kamaldeep, Sushma G Kaisar 1 , Shriram Tervankar 1 , Pravind Kumar 1 , Chandan Karekar 1 , UN Nayak 1 , NS Baghel 1 , MGR Rajan 1 , RK Gopalakrishnan

Radiation Safety System Division, 1 Radiation Medicine Centre, BARC, Mumbai, Maharashtra, India

Objective: To check the shielding adequacy radiometry was carried out in the facility housing newly installed PET-CT machine. The acceptable radiation level in occupied area at the working place, as per AERB regulatory guidelines is 1 μSv/h. Hence a stringent control on the shielding is mandatory. The inadequacy in shielding observed during radiation survey, was also corrected before commissioning of the system. Materials and Methods: To check the shielding integrity of the room, with the wall dimension of 228.6 mm concrete, doors with 4 mm lead lining and window having lead glass of 3 mm lead equivalent, the side walls were marked with points at a distance of 812.8 mm. The height of the points from floor was kept 812.8 mm (height of the gantry), doors and lead window were also marked to check the radiation streaming. 18F source of strength 632.33 MBq was used in PET gantry and 120 KV X-ray, 375 mA tube current and PMMA phantom were used in CT gantry. Radiation survey in the pre marked points and areas above and below the PET-CT room were carried out using plastic scintillator detector based instrument (sensitivity for 137Cs is 70 cps/μSv/h). Results: During radiometry, shielding losses at certain locations were observed. After rectification of shield, radiometry was repeated. The general radiation field observed in the grids for PET gantry was in the range of 0.06-0.41 μSv/h and in CT gantry, the range was 0.32-4.5 μSv/h. Conclusion: Radiometry of medical equipment is of utmost importance. An accurate radiometry will ensure the potential unwarranted exposure.


The International Nuclear and Radiological Event Scale for unplanned events affecting patients undergoing a medical procedure

Pankaj Tandon

Radiological Safety Division, Atomic Energy Regulatory Board, Niyamak Bhavan, Anushaktinagar, Mumbai, Maharashtra, India

Objective: The International Nuclear And Radiological Event Scale (INES) was developed with the aim of communicating the safety significance of events at nuclear installations in 1990. Since then, INES has been expanded to meet the growing need for communication on the significance of any event giving rise to radiation risks. Currently, INES covers a wide spectrum of practices, including use of radiation sources in industrial, in hospital, nuclear facilities and transport of radioactive material but the same can not cover the actual or potential consequences in patients exposed as part of medical procedure. Materials and Methods: Events are classified on the scale of seven levels: Level 4-7 are termed "accident" and level 1-3 "incidents". Events without safety significance are classified "Below scale/level 0". Events that have no safety relevance with respect to radiation or nuclear safety are not classified on the scale. Events are considered in terms of their impact on three different areas: Impact on people and the environment; impact on radiological controls at facilities; and impact on defence in depth. The impact on people and the environment can be localized, i.e. radiation doses to one or a few people close to the location of the event, or widespread as in the release of radioactive material from an installation. The impact on radiological barriers and control at facilities is only relevant to facilities handling major quantities of radioactive material such as power reactors, reprocessing facilities or large source production facilities. It covers events such as a reactor core met and spillage of significant quantities of radioactive material resulting from failures of radiological barriers, thereby threatening the safety of people and the environment. Reduction in defence in depth covers those events without impact on people, the environment or the facility but where the measures put in place to prevent accidents did not operate as intended. Now, International Atomic Energy Agency (IAEA) is in the process of finalizing a technical document on the use of INES scale for unplanned events affecting patients undergoing a medical procedure. This document will be helpful for each country during rating the events as and when it occurs. In some countries, there is already an established system for reporting the events affecting patients undergoing a medical procedure; this may be in diagnostic radiology, nuclear medicine and in radiotherapy practices. In India, every institute handling radionuclides or using radiation generating equipments have to periodically submit annual safety status report (ASSR) to the national regulatory body; atomic energy regulatory board (AERB). In the report, the institute has to give the details of any unusual occurrences occurred during the past 1 year. These unusual occurrences are mainly related to incidents related loss of radioactive sources or spillage of radioactive materials etc., but do not give the incidents related to patients. However, the IAEA document that is under final stage of preparation may recommend the member states to establish a system for rating the event. Conclusion: The main objective of bringing this document is to improve the radiation safety culture so that each institute has prevention and corrective action program in place. Regulatory body of India may also be making the format for reporting these incidents so that the mistakes which commonly occur during the diagnosis and treatment of patient can be avoided and improve the work culture for the benefit of the patient.


An approach for optimal shielding evaluation of the iodine-131 high dose therapy facilities

Jolly Joseph, Pankaj Tandon, Namitha Krishnakumar, Ashish Ramteke

Atomic Energy Regulatory Board, Mumbai, Maharashtra, India

Objective: The objective of this study is to evolve an approach for optimal shielding evaluation of the iodine-131 high dose therapy facilities. The general approach of shielding evaluation of I-131 high dose therapy facility is done by using the dose rate constant of I-131. The shortcoming of this method is that, other than the activity administered and distance of measurement, it does not consider some of the important factors which directly affect the radiation level. In this paper, an optimal shielding evaluation approach has been evolved which considers all the factors that affects the amount of shielding required. Materials and Methods: Dose rate constant for I-131 of 0.059 μSv m 2 /h MBq is generally being used for shielding evaluation. The patient's body absorbs a portion of the radiation coming out, resulting in its attenuation and this factor, called as attenuation factor μa tt , may be considered while evaluation. Radioactive decay of I-131 is another important factor for consideration during the evaluation. In most of the cases patients stay is between 24 and 72 h in order to adhere with the regulatory requirement. The Radioactive decay factor "R t" is used in this paper to include the effect of decay (physical and biological) for the shielding evaluation. Radioactive decay factor, R t = 1.443× (T eff /t) × (1- exp [−0.693 t/T eff ]) (A). Where "t" is the time duration for which the radiation level is required to be calculated and T eff is the effective half-life. The total dose "D t", received at a point which is at a distance of d meters for a time duration of "t" can be calculated as; D t = s0.059 (μSv m 2 /h MBq) × Ao (MBq) × t (h) × R t × μ att /d (m 2 ) (B). The total dose "D t" calculated shall be less than the dose limits stipulated by the regulatory body. The minimum distance to the occupied area from a shielded wall shall be considered. Transmission factor, B = 0.059 (μSv m 2 /h MBq) × Ao (MBq) × t (h) × R T × μ att × T/P d (m 2 ) (C). Where, "P0" is the permissible dose limit as stipulated by the regulatory body and "T" is the occupancy factor. The number of tenth value layer (TVL) of the shielding material required, n = −log (B). Appropriate wall materials like concrete, brick, lead, steel etc. can be used as deemed fit for the shielding purpose. Conclusion: An optimal shielding evaluation approach for the I-131 high dose therapy facility has been evolved which considers all the factors that affects the amount of shielding required. But, during the evaluation worst case scenario for radiation safety is also given importance.


Validation of sterility testing of radiopharmaceuticals by membrane filtration for reduction of the radiation dose

Arpit Mitra, Sangita Lad, Sushma Kaisar, Savita Kulkarni, MGR Rajan

MCF, RMC, BRIT/BARC, Parel, Mumbai, Maharashtra, India

Objective: USP/BP/EP/IP recommended sterility testing is performed by majority of the radiopharmacy centres, by membrane filtration or direct inoculation method depending upon the volume of radiopharmaceuticals. Since, direct inoculation method subject the analyst to high radiation exposure, present study was planned for standardization and validation of a low cost sterility testing of different radiopharmaceuticals by membrane filtration method. Materials and Methods: The membrane filtration method was initially validated by performing growth promotion test with Staphylococcus aureus, Bacillus subtilis Scientific Name Search  and Candida albicans at different cell concentrations ranging from 0 to 100 CFU and by serial dilution and spread plate method. The cell concentrations were also confirmed spectrophotometrically by taking OD at 560 nm. Radiopharmaceuticals like {18F}FDG, ( 18 F)NaF,{18 F}FLT, ( 18 F)FMISO, ( 177 Lu)DOTANOC and ( 99m Tc)TcO4− (1.5 ml) at concentrations of 3, 1 and 0.5 mci/ml, were aspirated in syringe and passed through 0.22 μm sterile syringe filter (polyether sulfone/mixed cellulose ester -membrane). This was followed by aspiration of 20 ml growth medium through the same filter into sterile syringe and incubation of syringe in vertical position at appropriate temperature for 14 days. For comparison, sterility testing (positive and negative cultures) was also performed for the same samples of radiopharmaceuticals by direct inoculation using 20 ml of growth medium. Results: Luxuriant growth same as the positive control, was observed in the all the radiopharmaceuticals and at various concentrations for all the micro-organisms used. This was confirmed by CFU using spread plate method. Positive and negative controls used in both the methods exhibited expected results of presence and absence of growth respectively. When sterility testing of (F-18) radiopharmaceuticals was performed by present method, analyst received dose of 40-300 μsv/h compared to the dose of 0.8-4.5 msv/h in the direct inoculation. Further, for [ 177 Lu]DOTANOC and [ 99m Tc]TcO4− for present method, 8-50 μsv/h dose was received by analyst compared to dose of 100-800 μsv/h in direct inoculation. All the doses were without any lead shielding. Conclusions: Sterility testing by present method was successfully validated for different radiopharmaceuticals and it decreased the radiation dose received by the analyst by 16-20 fold as compared to direct inoculation method.


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