Investigation of Workplace-like Calibration Fields via a Deuterium-Tritium (D-T) Neutron Generator.

Radiation survey meters and personal dosimeters are typically calibrated in reference neutron fields based on conventional radionuclide sources, such as americium-beryllium (Am-Be) or californium-252 (Cf), either unmodified or heavy-water moderated. However, these calibration neutron fields differ significantly from the workplace fields in which most of these survey meters and dosimeters are being used. Although some detectors are designed to yield an approximately dose-equivalent response over a particular neutron energy range, the response of other detectors is highly dependent upon neutron energy. This, in turn, can result in significant over- or underestimation of the intensity of neutron radiation and/or personal dose equivalent determined in the work environment. The use of simulated workplace neutron calibration fields that more closely match those present at the workplace could improve the accuracy of worker, and workplace, neutron dose assessment. This work provides an overview of the neutron fields found around nuclear power reactors and interim spent fuel storage installations based on available data. The feasibility of producing workplace-like calibration fields in an existing calibration facility has been investigated via Monte Carlo simulations. Several moderating assembly configurations, paired with a neutron generator using the deuterium tritium (D-T) fusion reaction, were explored.

Optimization of Labeling PSMAHBED with Ethanol-Postprocessed 68Ga and Its Quality Control Systems.

Radiolabeling of the prostate-specific membrane antigen (PSMA) inhibitor Glu-NH-CO-NH-Lys(Ahx) using the 68Ga chelator HBED-CC (PSMAHBED) allows imaging of prostate cancer lesions because of high expression of PSMA in prostate carcinoma cells and in bone metastases and lymph nodes related to the disease. The aim of this work was to optimize labeling of 68Ga-PSMAHBED using the efficient cation-exchange postprocessing of 68Ga as well as the development of a thin-layer chromatography (TLC)-based quality control system. Methods: Labeling was optimized for online ethanol-postprocessed 68Ga eluate investigating various parameters, such as buffer molarity (0.1-1 M), temperature (25°C-90°C), tracer amount (0.11-0.74 nmol), and labeling time. In addition, purification of the crude product was tested. For radio-TLC quality control, various mobile phases were analyzed using silica gel 60 plates and the results were validated using high-performance liquid chromatography. The most superior mobile phases were also applied on instant thin-layer chromatography (ITLC) silica gel plates. Results: Using optimized conditions, labeling yields of more than 95% were obtained within 10 min when ethanol-based postprocessing was applied using PSMAHBED amounts as low as 0.1 nmol. A higher precursor concentration (0.7 nmol) further increased labeling and quantitative yields to more than 98% within 5 min. In clinical routine, patient batches (>200 applications) with radiochemical purity greater than 98% and specific activities of 326 ± 20 MBq/nmol are obtained reproducibly. When TLC quality control was performed on silica gel 60 plates, 4 mobile phases with suitable separation properties and complementary Rf values were identified. Two systems showed equivalent separation on ITLC silica gel plates, with ITLC analysis finished within 5 min, in contrast to 20 min for the TLC system. Labeling of PSMAHBED was optimized for cation-exchange postprocessing methods, ensuring almost quantitative labeling and high nuclide purity of final 68Ga-PSMAHBED, making subsequent purification steps unnecessary. Conclusion: The new radio-TLC method allows quality control in a short time using a fast, reliable, low-cost method with little equipment complexity. Using this approach, the synthesis is easily adopted by automated synthesis modules.

Comprehensive Quality Control of the ITG 68Ge/68Ga Generator and Synthesis of 68Ga-DOTATOC and 68Ga-PSMA-HBED-CC for Clinical Imaging.

UNLABELLED: A good-manufacturing-practices (GMP) (68)Ge/(68)Ga generator that uses modified dodecyl-3,4,5-trihydroxybenzoate hydrophobically bound to a octadecyl silica resin (C-18) as an adsorbent has been developed that allows for dilute HCl (0.05N) to efficiently elute metal-impurity-free (68)Ga(3+) ready for peptide labeling. We characterized the performance of this generator system over a year in conjunction with the production of (68)Ga-labeled DOTATOC and Glu-NH-CO-NH-Lys(Ahx)-HBED-CC (PSMA-HBED-CC) intended for clinical studies and established protocols for batch release. METHODS: A 2,040-MBq self-shielded (68)Ge/(68)Ga generator provided metal-free (68)GaCl3 ready for peptide labeling in the fluidic labeling module after elution with 4 mL of 0.05N HCl. The compact system was readily housed in a laminar flow cabinet allowing an ISO class-5 environment. (68)Ga labeling of peptides using GMP kits was performed in 15-20 min, and the total production time was 45-50 min. Batch release quality control specifications were established to meet investigational new drug submission and institutional review board approval standards. RESULTS: Over a period of 12 mo, (68)Ga elution yields from the generator averaged 80% (range, 72.0%-95.1%), and (68)Ge breakthrough was less than 0.006%, initially decreasing with time to 0.001% (expressed as percentage of (68)Ge activity present in the generator at the time of elution), a unique characteristic of this generator. The radiochemical purity of both (68)Ga-DOTATOC and (68)Ga-PSMA-HBED-CC determined by high-performance liquid chromatography analysis was greater than 98%, with a minimum specific activity of 12.6 and 42 GBq/µmol, respectively. The radionuclidic ((68)Ge) impurity was 0.00001% or less (under the detection limit). Final sterile, pyrogen-free formulation was provided in physiologic saline with 5%-7% ethanol. CONCLUSION: The GMP-certified (68)Ge/(68)Ga generator system was studied for a year. The generator system is contained within the fluidic labeling module, and it is compact, self-shielded, and easy to operate using simple manual techniques. The system provides radiolabeled peptides with high (>98%) radiochemical purity and greater than 80% radiochemical yield. The (68)Ge levels in the final drug products were under the detection limits at all times. (68)Ga-DOTATOC and (68)Ga-PSMA-HBED-CC investigational radiopharmaceuticals are currently being studied clinically under investigational new drug (IND) applications submitted to the U.S. Food and Drug Administration.

Detailed evaluation of different (68)Ge/(68)Ga generators: an attempt toward achieving efficient (68)Ga radiopharmacy.

The present study is aimed at carrying out a comparative performance evaluation of different types of (68)Ge/(68)Ga generators to identify the best choice for use in (68)Ga-radiopharmacy. Over the 1 year period of evaluation, the elution yields from the CeO2-based and SiO2-based (68)Ge/(68) Ga generators remained almost consistent, in contrast to the sharp decrease observed in the elution yields from TiO2 and SnO2-based generators. The level of (68)Ge impurity in (68)Ga eluates from the CeO2 and SiO2-based (68)Ge/(68)Ga generator was always <10(-3)%, while this level increased from 10(-3)% to 10(-1)% in case of TiO2 and SnO2-based generators. The level of chemical impurities in (68)Ga eluates from CeO2 and SiO2-based (68)Ge/(68)Ga generators was negligibly low (<0.1 ppm) in contrast to the significantly higher level (1-20 ppm) of such impurities in eluates from other two generators. As demonstrated by radiolabeling studies carried out using DOTA-coupled dimeric cyclic RGD peptide derivative (DOTA-RGD2), CeO2-PAN and SiO2-based generators are directly amenable for radiopharmaceutical preparation, whereas the other generators can be only used after post-elution purification of (68)Ga eluates. Clinically relevant dose of (68)Ga-DOTA-RGD2 was prepared in a hospital radiopharmacy for non-invasive visualization of tumors in breast cancer patients using positron emission tomography.

Survey on radiation safety management (RSM) among Korean radiation workers who operate radiation generators or handle radioactive isotopes.

This study examined the knowledge, attitudes, and behaviors of radiation workers to radiation safety management (RSM) using survey questionnaires. Radiation workers are those who handle radiation generators, radioactive isotopes, and other radioactive materials for industrial uses. The survey was distributed to 861 radiation workers between 1 August to 5 September 2011. A knowledge of, awareness of, attitude toward, and behaviors related to RSM were analyzed by comparing the means and standard deviations. Both the knowledge and awareness of RSM among radiation workers were high. Although all questions about the awareness of RSM were answered correctly, there were also many negative responses. All questions regarding the attitude of radiation workers toward RSM were answered correctly, and their attitude toward and awareness of RSM were high. Overall, the results demonstrated that safety management is not taken seriously in many cases, highlighting the need for proper education in the future to raise awareness among radiation workers. Furthermore, it is important to establish a foundation for the efficient use of radiation based on the continuous management of radiation workers.

Quantitative online isolation of 68Ge from 68Ge/68Ga generator eluates for purification and immediate quality control of breakthrough.

The breakthrough of 68Ge from a 68Ge/68Ga-generator is one of the most sensitive parameters in the context of the clinical application of 68Ga-radiopharmaceuticals. The difficulty in its determination lies in the "spectroscopic invisibility" of 68Ge within an excess of 68Ga. The introduced method for determining the 68Ge content of the 68Ge/68Ga-generator eluate involves the quantitative separation of 68Ga from 68Ge, using a cation-exchanger. The eluate contains 68Ga free of 68Ge, which can be determined immediately, i.e. prior to the application of the 68Ga-radiopharmaceutical.

Recoil effect on beta-decaying in vivo generators, interpreted for (103)Pd/(103m)Rh.

The use of Auger emitters as potential radiopharmaceuticals is being increasingly investigated. One of the radionuclides of interest is (103m)Rh, which can be produced from (103)Ru or (103)Pd in an in vivo generator. A potential problem, however, is the recoil of the (103m)Rh out of the carrier molecule and even out of the target cell. In order to determine the likelihood of this happening in the (103)Pd/(103m)Rh, case calculations were made to prove that this does not happen. The equations were generalised for all radionuclides with an atomic mass of 10-240 as a tool for determining the recoil threshold of any beta-emitting radionuclide.

An improved 62Zn/62Cu generator based on a cation exchanger and its fully remote-controlled preparation for clinical use.

OBJECTIVE: The goal of this study was to develop an improved (62)Zn/(62)Cu generator based on cation exchange resin and remote preparation at high radioactivity scale for clinical use. METHODS: A natural Cu target was irradiated with proton beam in the energy range of 30-->19 MeV at a beam current of 10 muA for 1 h to obtain around 1.7 GBq of (62)Zn. The (62)Zn was isolated from the Cu target on an anion exchange column with more than 97% yield within 2.5 h from the EOB. The (62)Zn/(62)Cu generator was prepared by loading the (62)Zn(2+) on a Sep-Pak plus CM cartridge. RESULTS: The generator showed high elution efficiency ( approximately 96%) using a small volume (ca. 3 ml) of a 200-mM glycine solution with a very low breakthrough of (62)Zn (<0.1%). CONCLUSIONS: This (62)Zn/(62)Cu generator has been proven to be highly useful as a source of (62)Cu for the synthesis of (62)Cu-labeled compounds. The clinical application of [(62)Cu]Cu-ATSM produced with this generator has been already approved by the Institutional Review Board at the National Institute of Radiological Sciences.

Quality control of 99Mo/99Tcm generators: results of a survey of the Radiopharmacy Working Group of the Italian Association of Nuclear Medicine (AIMN).

A multicentre survey of the quality control of 99Tcm generators has been completed: 245 generators from seven different commercial sources were tested over a period of 2 years. The results indicate that the mean pH of the eluates was 5.8 +/- 0.6; the aluminium contents were typically < 10 ppm; the radiochemical purity was 99.8 +/- 0.4% and the median 99Mo content was 3.8 x 10(-4) percent. The elution profiles gave a volume of 1.9 ml to obtain 50% of the total eluted activity and of 4.9 ml to obtain 95%. Other radionuclide impurities and heavy metal breakthrough were evaluated by graphite furnace absorption spectrometry and inductively coupled plasma mass spectrometry. National guidelines for the standardization of radiopharmacy procedures are currently being compiled.

[A solid-state reference sample of iodine radionuclide activity for checking radiometers and a method for its production]. / Tverdotel'nyi standartnyi obrazets aktivnosti radionuklida ioda dlia poverki radiometrov i sposob ego izgotovleniia.

This development relates to nuclear engineering, concerns the sources of ionizing radiation of iodine radionuclides and the ways of their manufacture. The iodine radionuclides are used for calibration of radiometers that measure the specific activity of iodine radionuclide in biological specimens, in particular for their use to identify noncancer diseases. The purpose of the paper is to design a routine activity sample for iodine radionuclide in the solid-state source for transfer of an iodine radionuclide activity unit from the reference iodine radionuclide activity unit to an operating meter with a simultaneous decrease in testing errors and an increase, in terms of iodine radionuclide half-life, in the life of the reference sample by stabilizing the activity unit of iodine radionuclide (in terms of its half-life) by eliminating the process of sublimation of iodine radionuclide from the working surface of the source. The objective is achieved by the fact that the radionuclide is replaced on the lower internal build-up part of the container, fixed with the upper internal build-up part, then kept for at least 10-12 hours at least at 18-20 degrees C, packed into a hollow container and sealed with epoxide resin. According to the proposed technical design, reference activity samples of iodine radionuclide in the solid-state radiometric source with a preset activity value have been prepared. The testing error for iodine radionuclide activity is minimal and equal to 1.0-3.0% at a 0.95 confidence.

An assessment of factors which influence the effectiveness of the modified in vivo technetium-99m-erythrocyte labeling technique in clinical use.

This study assessed factors which may contribute to suboptimal image quality when the modified in vivo erythrocyte labeling technique is used with standard clinical 99mTc activities. For each assessment duplicate or triplicate blood specimens were withdrawn from > or = 10 patients, into syringes containing 700-900 MBq 99mTc as pertechnetate. After incubation the percent of 99mTc which was not bound to erythrocytes at blood re-injection time (%Unbound 99mTc), was measured and compared when one of four factors was varied. The most significant results, in descending order of measured effect were: [table: see text] Our data suggest that the requirements for optimal erythrocyte labeling with standard clinical 99mTc activities are: (A) Erythrocyte tinning time between 10 and 30 min; (B) blood volume > or = 3 ml; (C) blood incubation time > or = 20 min; and (D) Generator ingrowth time < or = 24 hr.

Comparison of four commercially available 81Rb-81mKr generators: quality testing using a breathing lung phantom.

Four commercially available 81Rb-81mKr generators, including their elution systems, were compared. Considerable differences in useful output between the various generators were found. When a standardized elution system was used, the differences became much less. Differences in the tubing and inaccuracy of the flow meters were found to be the major causes of the original differences. The user should be aware of a loss of output that may occur when the generator is eluted with dry air.

Fission 99Mo/99mTc Generators--a study of their quality.

Quality was controlled of seven different 99Mo/99mTc generators sold in the Federal Republic of Germany, which were all loaded with fission molybdenum. Aim of the work was to test the 99mTc eluates for their radionuclidic and radiochemical purity. Furthermore, the 99Mo content of the aluminium oxide column was measured. Determination of the elution efficiency, measurement of the pH-value of the eluate as well as a test to detect soluble aluminium in the eluate were also performed. In addition to the chemical and physical examinations, application, quality of accessories and their practicability were tested at all generators. The methods of analysis chosen were, among others, gamma spectrometry and thin-layer chromatography; the measuring methods were a site-sensitive proportional counter (measurement of the distribution of radioactivity of the chromatograms) and atomic absorption spectrometry (Al content of the eluates). The quality control of the 99mTc eluates had satisfying results. The eluates showed - with one exception - high and sufficient radionuclidic purity and very good radiochemical purity. A considerable overload of the columns with 99Mo at the time of reference was not found. The elution yields with values between 85 and 122% were in good agreement with the requirements. All eluates had pH-values between 5.0 and 6.5, and an aluminium content below 1 microgram/ml. The generators had good performance and proved generally to be a reliable source of 99mTc-pertechnetate. The application was safe and, with some exceptions, fulfilled the requirements.

Radiation dose to various age groups from radionuclide impurities in 99mTc-pertechnetate (fission product 99Mo generator) radiopharmaceutical preparations.

Radionuclide impurities of radiopharmaceuticals do not provide any diagnostic information, but contribute to the radiation dose to the patient. For 99mTc, even small amounts of long lived radionuclide impurities may contribute significantly to the dose from 99mTc itself. In this preliminary study, estimates have been made for the radiation dose to specific organs and for the effective dose equivalent from impurities in 99mTc-sodium pertechnetate prepared from fission product 99Mo. In addition to 99Mo, 131I, 103Ru, 89Sr and 90Sr at maximum permissible levels as stipulated in the European Pharmacopoeia, 110mAg and 239Pu were taken as representative nuclides for beta-gamma emitters and alpha emitters. Using all of these radionuclides at such maximum permissible levels of impurity as stated in the European Pharmacopoeia, the total contribution from the impurities to the effective dose equivalent is approximately 30% of the effective dose equivalent from 99mTc-pertechnetate. In terms of specific organs, the liver and kidneys can receive doses of 3-4 times and 1-3 times the dose from 99mTc-pertechnetate itself. Although these are worst case calculations, a consistent approach should be reconsidered to limit the additional effective dose equivalent from impurities to e.g. 10% of the effective dose equivalent from 99mTc itself.

A breathing lung phantom for 81mKr lung ventilation studies its use in dosimetry and quality control.

The construction of a breathing lung phantom that can be used to measure the amount of radioactive gas in the lungs as well as to determine the absorbed dose is described. For a lung ventilation study that consists of 6 views of 300 kcnts each, an effective dose equivalent of 50 microSv was calculated. The phantom is also suitable for comparison of different generator systems.

[A new sterile generator of Tc 99m. Quality control]. / Un nuovo generatore sterile di 99mTc. Controlli di qualità.

Production has started on a new Tc 99m sterile generator in Italy. We give the results of a series of quality controls in order to have a first evaluation of the product. The elution profile and the volume, activity, pH, aluminium concentration, 99 Mo content and radiochemical purity of the eluate were examined. Radiochemical purity tests were also carried out with some of the more widely used "in vivo" radiopharmaceuticals. The results are in line with the Official Pharmacopeia and good technical standards.