Year : 2011 | Volume
: 26 | Issue : 5 | Page : 34--36
|How to cite this article:|
. Radiation Safety.Indian J Nucl Med 2011;26:34-36
|How to cite this URL:|
. Radiation Safety. Indian J Nucl Med [serial online] 2011 [cited 2020 Nov 30 ];26:34-36
Available from: https://www.ijnm.in/text.asp?2011/26/5/34/90732
Pediatric dosimetry of 18 F-FDG whole body PET/CT scans
Aruna Kaushik, Anupam Mondal 1 , Rajnish Sharma 1 , Maria D'Souza 1 , Abhinav Jaimini 1 , Dinesh Singh 1 , Anjani Tiwari, AK Mishra, BS Dwarakanath 2
Division of Cyclotron and Radiopharmaceutical Sciences, 1 Division of PET Imaging, 2 Division of Radiation Biosciences, Institute of Nuclear Medicine and Allied Sciences, Brig. S.K. Mazumdar Marg, Timarpur, Delhi, India
Introduction: A combined 18 F-FDG (18 F-2-deoxy-D-glucose) positron emission tomography/ computed tomography (PET/CT) scan provides both the metabolic information from FDG-PET and anatomic information from CT in a single examination. The use of PET/CT for management of malignancies in children has increased over the past few years. This raises an important consideration of radiation exposure in children since they are relatively more radiosensitive than adults and also have a potential for a longer life thereby increasing the probability of manifestation of late radiation effects; particularly cancer. Unfortunately, the data regarding the doses received by children from undergoing such examinations is scarce. The present study aims at estimating the effective doses to paediatric patients from whole body 18 F-FDG PET/CT studies. Objective: The purpose of the study is to estimate the radiation doses to children from undergoing whole body PET/CT scans using 18 F-FDG. Materials and Methods, Subjects: A random sample of hundred (age range: 0-2, 2-6, 6-12 and 12-18 years) paediatric patients from 18 F-FDG PET/CT studies was included for dosimetry estimates. The average weight (in kg) of the children for the above mentioned age range was 10.3±2.1, 16.3±4.1, 29.6±8.8, 42.2±10.7 and height (in cms) was 87±8.9, 104.2±10.5, 130.6±14.6, 155.5±10. The average amount of activity (in MBq) of 18 F-FDG injected for the different age ranges was 150±36, 199.15±45.8, 246.7±49.7, and 288.2±49.1 respectively. The use of patient's data for the purpose of estimating radiation doses was approved by the institutional human ethics committee. Protocol for whole body PET/CT scan: The protocol for whole body 18 F-FDG PET/CT examinations comprised of (a) a topogram, or scout scan, for positioning; (b) a spiral CT scan for attenuation correction; (c) a PET scan over the same axial range as the CT scan. The patients were scanned from the top of the skull to mid thigh. Internal dosimetry: The effective doses are computed by using the dose coefficients and tissue weighting factors recommended by ICRP. Effective doses, E were estimated by:
E = A • ΓE
A is the administered activity and ΓE is the dose coefficient for the effective dose.
External dosimetry (CT): The effective dose from the CT component was computed using CT-Expo software on the basis of the scan parameters and the characteristics of the model of CT system, Lightspeed 16 stored in the software data base. The scan parameters used in the spiral CT scan of whole-body 18 F-FDG PET/CT studies were 120 kV, 110 mA, 0.8 s scan duration, beam width 10 mm, reconstructed slice thickness 3.75 mm, Table feed 17.5 mm and pitch 1.75:1. Results and Conclusion: The total effective dose among the paediatric patients of the combined PET/CT studies, calculated by summing the effective doses of PET and CT scanning from 18 F-FDG whole body PET/CT examination was 24.75±3.4 mSv, 20.05±2.6 mSv, 18.3±1.8 mSv and 16.3±1.2 mSv for the age ranges 0-2, 2-6, 6-12 and 12-18 years respectively. The dose received by the children in the age group 12-18 years is comparable to the adult doses reported by various studies. However, the dose received by the children in the age range 0-2 years is 34%, 2-6 years is 18% and 6-12 years is 10.9% more than the dose received by children in the age range 12-18 years. This may be due to higher metabolic rate, higher radiosensitivity and relatively more activity administered per unit mass of the body to the paediatric patients in the lower age groups. Higher effective dose to children in these age groups results in relatively higher risk of radiation as compared to adults. However, the effective dose and in turn the radiation risk to paediatric patients can be reduced by optimising the protocol for PET/CT scanning that involves administering the amount of PET pharmaceutical as per their body weight and also modulating the CT scan parameters according to the size and weight of the patient.
Radiological safety of bone mineral densitometry equipment
Santosh Kumar, Aruna Kaushik, Maria D'Souza, Dinesh Singh, Rajnish Sharma, Anupam Mondal
Division of PET Imaging, Institute of Nuclear Medicine and Allied Sciences, Brig. S.K. Mazumdar Marg, Timarpur, Delhi, India
Introduction: Osteoporosis is a highly prevalent disease leading to increased risk of bone fractures. Bone mineral densitometry (BMD) is a well accepted clinical tool for the diagnosis and management of osteoporosis. There are several different modalities for BMD such as Dual Energy X-ray Absorptiometry, Quantitative Ultrasound, Radiographic Absortiometry and Quantitative Computerized Tomography. Measurement of bone mineral density (BMD) by dual energy X-ray absorptiometry (DXA) is now well established as the method of choice for osteoporosis assessment. Objective: This study was conducted to assess the radiation dose to patients and staff from the standard scan modes. Materials and Methods: We have conducted a survey to determine the whole body radiation dose received specifically through operation of the dual energy X-ray absorptiometry (DXA) system Lunar Prodigy bone densitometer, M/s GE Healhcare installed in our institute. This is operated by departmental technologist staff working in rotation. Each patient routinely receives a standard DXA scan of the whole body, spine, hip region and wrist. The scan parameters are 76 kV, 0.15 mA for the whole body, 76 kV, 3mA for spine, 76 kV, 3 mA for the hip region and 76 kV, 0.15 mA for the wrist. The radiation levels were measured at various room locations [Figure 1]. The patient doses were observed from the values displayed on the Lunar Prodigy console for each scan. Results: The radiation levels measured at various locations in the room housing the BMD system are tabulated in [Table 1]. The gamma radiation level measured at the locations of interest was less than the permissible limit of 10 μSv/h for working hours. The dose to patients from the four scan modes are tabulated in [Table 2]. The doses to the patients are much below the standard radiographic examinations. Conclusion: The radiation levels at various locations in the room housing the equipment are much below the permissible levels. The equipment is safe from radiation safety standpoint. However, receiving unnecessary exposure by standing near the couch should be avoided and ALARA (As low as Reasonably Achievable) principle should be followed. The radiation dose received by patients and the risk of cancer from BMD is much lower as compared to other radiographic procedures.
Radiopharmacy contamination in nuclear medicine - A survey report
Anchal Ghai, Somnath Mardi, Pradeep Kumar, Sarika, BR Mittal, B Singh
Department of Nuclear Medicine, PGIMER, Chandigarh, India
Objectives: To conduct the radiation survey of the category IV Nuclear Medicine department and to further measure and compare the level of loose contamination in areas with high levels of exposure at two different time points (morning and evening) using wipe test. Materials and Methods: The radiation monitoring survey of thes Department of Nuclear Medicine was performed using G.M. counter based survey meter. The survey was performed in 12 rooms of the department. The room exposure levels were recorded in mR/hr at three different time intervals namely, before the start of day's work, during work and at the end of the day's work. Wipe testing was done in the areas showing high exposure levels twice a day, in morning and evening. Requirements for wipe testing were cotton swabs dipped in alcohol, forceps, a pair of gloves, test tubes with stand and well counter with calibrating sources. An area of 936.36 cm 2 was swiped for each wipe sample collected. The data was collected for 30 different days. Results: The average exposure in the non radioactive areas was found to be equal to background. The daily average exposure in the controlled areas of the department i.e hot lab and RIA lab were 0.174 mR/hr and 0.063 mR/hr respectively and that in the controlled supervised areas was almost equivalent to background. Wipe testing was performed at 7different locations in the hot lab (near fumehood, injection bed, L-bench 1, L-bench 2, sink, centre of lab and background). Injection bed, sink and area outside the door of radiopharmacy showed high contamination levels as compared to the recommended levels. Conclusion: As a part of surveillance program of Nuclear Medicine laboratories, daily radioactivity contamination monitoring using wipe test in the areas of concern especially in extremely busy nuclear medicine departments (with high patient's load) will provide 'guidance' for careful compliance with radioactivity handling practices.
Radiation exposure during leukocyte labelling process
Sarika, A Bhattacharya, R Kochhar 1 , BR Mittal
Department of Nuclear Medicine, 1 Department of Gastroenterology, Postgraduate Institute of Medical Education and Research, Chandigarh, India
Objective: Leukocyte labelled imaging is considered a gold standard technique for the diagnosis of the infection. However, the radiolabeling of the white blood cells is a time consuming process, which requires open handling of the radio-nuclides for a considerable period of time leading to high radiation exposure to the personnel. In this study an attempt was made to record the total exposure received and the total time taken for the leukocyte labelling using two different radio-nuclides. Materials and Methods: The average exposures on the workbench during the various steps of the radiolabeling were recorded using a survey meter. The personnel exposure was recorded using a pocket dosimeter. Results: The average total time taken for the leukocyte labelling using 99m-Tc HMPAO and 18 F-FDG were 23.08 and 35.42 minutes respectively. The mean radiation exposure received per radiolabeling procedure for 99m-Tc HMPAO and 18 F-FDG were 2.04 microSievert and 6.44 microSievert respectively. The average exposures on the workbench during the various steps of the radiolabeling are given in the [Table 1]. The time taken for the 18 F-FDG leukocyte labelling was higher than the Tc99m HMPAO leukocyte labelling due to longer incubation period required for the former process. The mean radiation exposure received is also higher in the 18 F-FDG leukocyte labelling process due to the higher gamma ray dose constant of the 18 F radionuclide. Conclusion: The radiolabeling of leukocytes by any radionuclide can be safely performed using the basic radiation safety principle of TDS. The study demonstrated the safe leukocyte radiolabeling with two different radio-nuclides using the same set of radiation shielding equipments. However the radiation exposure is relatively higher in the 18 F-FDG leukocyte labelling process than the 99m-Tc HMPAO leukocyte labelling process.
Radiological safety in using N-13 radioisotope in the PET facility in India
Ajay Kumar Gocher, Pankaj Tandon
Radiological Safety Division, Atomic Energy Regulatory Board, Mumabi, India
Introduction: Positron emission tomography (PET) is the most viable diagnostic tool presently available to diagnose the functional abnormality of the organ in the human body. The most commonly used PET radio pharmaceutical is FDG labelled with F-18 radioisotope having reasonably good half life (110 min), making it available to the PET installation. These PET installations are approved by Atomic Energy Regulatory Board (AERB) for calculated work load (pt/week) based on the shielding adequacy of the facility. In PET installation mostly work load is limited by uptake room because patients have to spend more time in uptake room than imaging room. For some other study (Myocardial perfusion study) Nuclear Medicine physician require to administer N-13 labelled radiopharmaceutical to the patient having half life of 10 min. Because of its short half life and requirement of study, activity of 370-1110 MBq needs to be administered within the imaging room. In this case the patient need not go to the uptake room. Therefore, there is a need to re- calculate the work load (pt/week) for imaging room for N-13 administered patient in the approved PET installation using F-18. Objective: To calculate the work load (pt/week) for N-13 administered patient in the PET-CT installation and verify, whether the workload for using N-13 is same as the workload for using F-18. Materials and Methods: Typical administered activity of N-13 labelled compound is around 370-1110 MBq and maximum scan time is 20 min including patient set-up on the couch. The work load has been calculated as per AAPM TG-108 report. Result and Conclusion: From the theoretical calculations it has been noted that the work load for using N-13 will not be same as using F-18. One may practice with N13 labelled compound in the installed PET-CT but keep in mind about the reduction in the work load by keeping the radiation level with in the permissible limit.
High radiation area survey during the running of the Cyclotron
Amandeep Kaur, Sarika Sharma, BR Mittal
Department of Nuclear Medicine and PET, PGIMER, Chandigarh, India
Introduction: Routine cyclotron survey is an important part of overall radiation safety in the cyclotron facility. This procedure is routinely performed by the Medical Physicist/RSO during the operation of the cyclotron. This helps to ensure the proper radiation safety and prevents unjustified radiation exposure to the working staff, radiation workers, patients and visitors. The present study was performed to check the areas with high radiation exposure during the running of the cyclotron. Materials and Methods: Ionization chamber based detector and Geiger Muller counter were used to record the exposure levels before, during and after operating the cyclotron. The readings were recorded at various locations where a high radiation field was expected. The results were recorded, tabulated and analyzed. Results: Highest exposure level (0.93μSv/hr) was found in the hot lab after the bombardment. The next highest exposure level was found near the vault wall (0.26μSv/hr) during production of the radioactivity in the cyclotron. The average exposure levels near the vaults door, console station, outer wall, back wall and adjacent rooms using the GM counter were 0.17, 0.16, 0.14, 0.18 and 0.14μSv/hr respectively. The average background levels recorded was 0.13μSv/hr. Conclusion: The radiation survelliance around the self shelded cyclotron reflected that the radiation exposure levels outside the shield are minimal. The study showed no leakage of the radiation outside the vault. All the values are below recommended levels of exposure.
Delay Tank- An overview in Indian regulatory context
Subrata Pathak, Pankaj Tandon
Radiological Safety Division, Atomic Energy Regulatory Board, Mumbai, India
Introduction: Delay tank is a tank or reservoir for the temporary holdup of radioactive waste coming out of a high dose radioiodine ward, to allow the activity to decay and bring down to the prescribed limit for discharge. Objective: The requirement of Delay tank in a high-dose radioiodine ward where patients suffering from Ca-thyroid and neuroendocrine tumour are being treated with Radio-Iodine (RI) [131I] and Lu177 has been reviewed in light of present Indian regulatory context. Materials and Methods: The clearance of RI from patients, its accumulation in the Delay tank and the amount of activity remaining in the Delay tank following physical decay has been simulated using simple mathematical model. Different case scenarios have been taken into account to achieve a better estimation of the cumulative activity in the Delay tank. Results and Conclusion: This theoretical calculation confirms the requirement of Delay tank in high dose radioiodine facilities to meet the present regulatory limits for discharge of radioactivity into a public sewerage line.