(1/472) MIRD Pamphlet No. 14 revised: A dynamic urinary bladder model for radiation dose calculations. Task Group of the MIRD Committee, Society of Nuclear Medicine.
The constant-volume urinary bladder model in the standard MIRD Pamphlet No. 5 (Revised) phantom has recognized limitations. Various investigators have developed detailed models incorporating more physiologically realistic features, such as expanding bladder contents and residual volume, and variable urinary input rate, initial volume and first void time. We have reviewed these published models and have developed a new model for calculation of radiation absorbed dose to the urinary bladder wall incorporating these aspects. METHODS: The model consists of a spherical source with variable volume to simulate the bladder contents and a wall represented by a spherical shell of constant volume. The wall thickness varies as the source expands or contracts. The model provides for variable urine entry rate (three different hydration states), initial bladder contents volume, residual volume and first void time. The voiding schedule includes an extended nighttime gap during which the urine entry rate is reduced to one-half the daytime rate. RESULTS: Radiation-absorbed dose estimates have been calculated for the bladder wall surface (including photon and electron components) and at several depths in the wall (electron component) for 2-18F-fluoro-2-deoxy-D-glucose, 99mTc-diethylenetriaminepentaacetic acid (DTPA), 99mTc-HEDP, 99mTc-pertechnetate, 99mTc-red blood cells (RBCs), 99mTc-glucoheptonate, 99mTc-mercaptoacetyltriglicine chelator (MAG3), 99mTc-methylene diphosphonate (MDP), 99mTc-hexamethylpropylene amine oxime (HMPAO), 99mTc-human serum albumin (HSA), 99mTc-MIBI (rest and stress), 123I-/124I-/131I-OIH, 123I/131I-NaI, 125I-iothalamate, 111In-DTPA and 89Sr-SrCl. CONCLUSION: The new model tends to give a higher radiation absorbed dose to the bladder wall surface than the previous models. Large initial bladder volumes and higher rates of urine flow into the bladder result in lower bladder wall dose. The optimal first voiding time is from 40 min to 3 hr postadministration, depending on radiopharmaceutical. The data as presented in tabular and graphic form for each compound provide guidance for establishing radiation absorbed dose reduction protocols. (+info)
(2/472) Practical aspects of radiation safety for using fluorine-18.
The use of positron-emitting nuclides is becoming routine in nuclear medicine departments today. Introducing these nuclides into the nuclear medicine department can be a smooth transition by instituting educational lectures, radiation safety protocols and patient education. The radiation safety concerns of the technical staff, physicians and ancillary personnel are important and must be addressed. Nuclear medicine departments can be optimistic about implementing PET imaging while staying well within ALARA guidelines. After reading this article, the technologist should be able to: (a) describe at least three ways to reduce the radiation dose to the technologist during the performance of PET imaging procedures with 18F; (b) discuss the relationships between gamma-ray energy, the amount of activity administered to a patient, exposure time and occupational dose; and (c) describe one strategy to minimize the radiation dose to the bladder in patients who have received 18F. (+info)
(3/472) Radiation exposure from gallium-67-citrate patients.
OBJECTIVE: Serial monitoring of patients was performed to determine the radiation exposure contributed by patients injected with 67Ga-citrate to their surroundings. Radiology and nursing staff distance exposure estimates were made for various patient care tasks and imaging tests. METHODS: Fifteen adult patients were surveyed early (mean 4.3 min) and 11 of the 15 were surveyed at 3 d (mean 68.8 h) postinjection. The standard adult lymphoma imaging activity of 333-407 MBq (9-11 mCi) resulted in a range of 3.7-8.1 MBq/kg (0.1-0.22 mCi/kg). Dose rate measurements were made in the anterior, posterior, and left and right lateral projections at the level of the umbilicus, at distances of patient's surface and at 30.5 cm and 100 cm with a calibrated ion chamber. Time of contact-routine task analyses also were obtained for nursing and radiology personnel. Using a radiation survey-derived biexponential pharmacokinetic relationship, radiation exposures were determined for hospital personnel and family members at various times after injection. RESULTS: Based on the study population survey results, the mean instantaneous exposures (microSv/h) for an administered activity of 370 MBq (10 mCi) 67Ga-citrate were determined. The task analyses revealed the maximum patient contact time for any procedure performed at a distance equal to, or less than, 30.5 cm was 30 min. CONCLUSION: The quantitation of radiation exposure scenarios from 67Ga-citrate patients has determined that no special precautions are necessary for medical personnel when performing routine tasks associated with these patients. (+info)
(4/472) Bremsstrahlung radiation exposure from pure beta-ray emitters.
With increasing therapeutic use of radionuclides that emit relatively high-energy (>1 MeV) beta-rays and the production in vivo of bremsstrahlung sufficient for external imaging, the potential external radiation hazard warrants evaluation. METHODS: The exposure from a patient administered beta-ray-emitting radionuclides has been calculated by extending the National Council on Radiation Protection and Measurement model of a point source in air to account for biologic elimination of activity, the probability of bremsstrahlung production in vivo and its mean energy and the absorption by the patient's body of the bremsstrahlung thus produced. To facilitate such calculations, a quantity called the "specific bremsstrahlung constant" (in C/kg-cm2/MBq-h), betaBr, was devised and calculated for several radionuclides. The specific bremsstrahlung constant is the bremsstrahlung exposure rate (in C/kg/h) in air at 1 cm from a 1 MBq beta-ray emitter of a specified maximum beta-ray energy and frequency of emission in a medium of a specified effective atomic number. RESULTS: For pure beta-ray emitters, the retained activities at which patients can be released from medical confinement (i.e., below which the effective dose equivalent at 1 m will not exceed the maximum recommended value of 0.5 cSv for infrequently exposed members of the general public) are extremely large: on the order of hundreds of thousands to millions of megabecquerels. CONCLUSION: Radionuclide therapy with pure beta-ray emitters, even high-energy beta-ray emitters emitted in bone, does not require medical confinement of patients for radiation protection. (+info)
(5/472) Internal radionuclide radiation dosimetry: a review of basic concepts and recent developments.
Internal dosimetry deals with the determination of the amount and the spatial and temporal distribution of radiation energy deposited in tissue by radionuclides within the body. Nuclear medicine has been largely a diagnostic specialty, and model-derived average organ dose estimates for risk assessment, the traditional application of the MIRD schema, have proven entirely adequate. However, to the extent that specific patients deviate kinetically and anatomically from the model used, such dose estimates will be inaccurate. With the increasing therapeutic application of internal radionuclides and the need for greater accuracy, radiation dosimetry in nuclear medicine is evolving from population- and organ-average to patient- and position-specific dose estimation. Beginning with the relevant quantities and units, this article reviews the historical methods and newly developed concepts and techniques to characterize radionuclide radiation doses. The latter include the 3 principal approaches to the calculation of macroscopic nonuniform dose distributions: dose point-kernel convolution, Monte Carlo simulation, and voxel S factors. Radiation dosimetry in "sensitive" populations, including pregnant women, nursing mothers, and children, also will be reviewed. (+info)
(6/472) A simple solution to prevent the loss of radioactive spot markers.
OBJECTIVE: Most nuclear medicine technologists have experienced the misplacing and/or the loss of a radioactive spot marker. We report on a simple solution to prevent or at least minimize the loss of radioactive spot markers. METHODS: One end of a metallic beaded chain was attached to the side of 57Co spot marker using repair putty. The other end of the beaded chain was attached to a lead shield that housed the radioactive source when not in use. RESULTS: This design has allowed easy, unobstructed use of the 57Co spot marker for marking the right or left side and anatomical position during imaging while preventing its loss. CONCLUSION: A radioactive spot marker that is attached to a lead shield by a beaded chain is a simple way to prevent its loss while allowing it to be used easily during imaging. (+info)
(7/472) Volatility of radiopharmacy-prepared sodium iodide-131 capsules.
OBJECTIVE: The aims of this study were to quantify the extent of volatilization from 131I-NaI therapeutic capsules prepared in a centralized radiopharmacy and to quantify the amount of volatile 131I released from a dispensing vial containing a compounded 131I-NaI therapy capsule. METHODS: Therapy capsules were prepared by injecting 131I oral solution into capsules containing anhydrous dibasic sodium phosphate. Volatilized activity was obtained by filtering air drawn across samples that were placed open on the bottom of a sample holder cup. Volatile 131I was captured by filtering it through 3 triethylenediamine-impregnated carbon cartridge filters, arranged in series. To quantify the amount of volatile 131I released from a dispensing vial during a simulated patient administration, a vial containing a compounded 131I therapy capsule was opened inside a collapsible plastic bag and all the air was drawn across TEDA-impregnated carbon cartridge filters. RESULTS: The 370-MBq (10-mCi) 131I capsules from the first part of the experiment released an average of 0.035% (SD 0.031%) of the capsule activity on the first day, 0.012% (SD 0.002%) on the second day, and 0.012% (SD < 0.001%) for days 3 through 5. The 37-MBq (1-mCi) 131I capsules released an average of 0.058% (SD 0.025%) on the first day, 0.029% (SD 0.009%) on the second day, and 0.020% (SD 0.004%) on the third day. The activity released from the vial during a simulated patient administration was 0.00093% of the 131I capsule activity. CONCLUSION: The amount of 131I, which volatilized daily from the exposed therapy capsules, was a small percentage of the capsule activity. The volatile 131I that would be released during a patient administration was much less than the activity that volatilized from the exposed therapy capsules. (+info)
(8/472) The relationship between elution time and eluate volume using the Ultra-TechneKow DTE technetium-99m generator.
OBJECTIVE: The new Ultra-TechneKow Dry Ship Top Elute 99mTc generator (UTK-DTE generator; Mallinckrodt Medical, Inc., St. Louis, MO) was devised to facilitate fractionated elution with an ergonomically designed elution shield. Fractionation is accomplished traditionally by visually observing the eluted volume through 2 layers of leaded glass windows located in a lighted elution shield and generator auxiliary shield. The goal of our study was to use elution time to determine the endpoint for obtaining the required volume of 99mTc-eluate from a UTK-DTE generator. METHODS: After triplicate elution at several predetermined elution times, the initial weight of the evacuated collecting vial was subtracted from the total weight after elution to determine the elution volume. RESULTS: A quadratic relationship was established between the eluate volume (v, mL) and elution time (t, s) (v = 0.3594 + 0.1889 t - 0.0009 t2). This equation is suitable for use with the 10-mL elution vial. This formula may not be accurate for the first elution since the UTK-DTE generator is a dry-column generator when shipped. The following elution times were calculated for some commonly eluted volumes: 2 mL (9 s), 4 mL (22 s), 5 mL (28 s), 7 mL (45 s), and 10 mL (88 s). CONCLUSION: Our calculated elution time method can be used to predict the eluate volume from a UTK-DTE generator. (+info)