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.

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.

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]

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.

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.

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 new sterile generator of Tc 99m. Quality control]

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.

Modified in vivo behaviour of 99mTc-labelled liver-spleen preparations due to 99Mo/99mTc-generator eluate impurities

At irregular intervals, a 99mTc-labelled microcolloid for liver and spleen scintigraphy was reported to accumulate in the lungs, in the absence of any pathophysiological condition usually associated with this phenomenon. These findings were in direct contrast to the results obtained during quality control assays performed on the production batches of the kit. The authors tested the hypothesis that the presence of Al3+ ions in the generators eluate used to label the kit caused the formation of larger size aggregates. The experimental results suggest that the 99mTc-pertechnetate solution used for labelling contained far more than 40 micrograms Al3+ per ml, hence a many-fold increase of the maximum level tolerated by the pharmacopoeia. The authors advise caution when interpreting findings of pulmonary activity in the absence of clear evidence of pathological conditions that are sometimes associated with this biodistribution pattern.

Radioassay problems associated with the clinical use of a 82Rb radionuclide generator

The short-lived positron emitting radionuclide 82Rb (t1/2 1.27 min) is conveniently available from a 82Sr/82Rb generator system. The parent nuclide (t1/2 25.5d) produced from the spallation of molybdenum, has associated with it varying quantities of other long-lived strontium radionuclides, namely 85Sr, 89Sr and 90Sr. It is important therefore in the clinical use of such material that the levels of strontium radionuclides being administered to patients is carefully assayed and controlled. The problems associated with these measurements are discussed with special reference to the radiation dose received by the patient and the problems in resolving overlapping peaks with different FWHMs.

Plutonium in [99mTc] Pertechnetate for clinical use: concise communication

Pertechnetate from 31 different Tc-99m generators was analyzed to determine if alpha-emitting radionuclides were present. Significant amounts of Pu(239 + 240) were found in the pertechnetate eluted from neutron-activated molybdenum distributed by one manufacturer. The Pu(239 + 240)/Tc-99m activity ratio in the first eluate from these generators varied between 1.3 . 10(-12) and 1.4 . 10(-11). The Pu(239 + 240) contamination is believed to originate from uranium impurities in the molybdenum batch. The build-up of Am-241, the daughter decay product of Pu-241, has also been demonstrated. The estimated Pu-241/Pu(239 + 240) activity ratio at the reference time was 8:1. It is strongly recommended that upper limits for the plutonium content of eluates from nonfission-produced molybdenum should be included in the various pharmacopoeias and that restrictions be placed on the uranium content of molybdenum used for the production of Mo-99 by neutron activation.

Results of regular study on radionuclidic purity of 99mTc obtained from 99Mo--99mTc generators

A total of 39 99mTc eluates obtained from 9 99Mo--99mTc generators delivered by The Radiochemical Centre Amersham during one year was studied with regard to their radionuclide purity. Using a Ge(Li) spectrometer the contaminants 60Co, 103Ru, 131I, 134Cs, 140Ba, 140La and 188Re were found in 99mTc-eluates with average levels ranging from 2.9 X 10(-3) to 2.8 X 10(-1) per cent of 99mTc activity. The additional total body absorbed dose caused by these contaminants, as calculated from their average content in 99mTc eluates, was less than 1% of the dose due to 99mTc pertechnetate.

Evaluation and application of alumina-based Rb-82 generators charged with high levels of Sr-82/85

Generator-produced Rb-82, a 75-sec positron emitter with potential for myocardial blood-flow imaging, was studied with various ion-exchange columns to evaluate the characteristics of alumina as an adsorber for the 25-day Sr-82 parent. Test columns of alumina, Bio Rex 70, and Chelex 100 were loaded with multimillicurie amounts of no-carrier-added Sr-82/Sr-85 (Sr-85 is a production contaminant). The breakthrough of Sr-82/Sr-85, and the yield of Rb-82, were determined for long-term elutions from each column with up to 4 liter of 2% NaCl solution at pH 8 to 9. The breakthrough of Sr-82/85 was 10(-6)-10(-5) from aluminal 10(-6)-10(-4) from Chelex 100 and Bio Rex 70. The effects of eluent flow rate and concentration, and of alumina volume, on the breakthrough and yield were also studied. An improved and automated Rb-82 generator was used for myocardial and brain blood-flow studies in experimental animals and in man; it was equipped with solenoid flow-control valves and five in. of lead shielding for the alumina columns, which were charged with 25-50 mCi Sr-82 (100-150 mCi Sr-85). The Rb-82 generator with alumina column provided up to 20-40 mCi of Rb-82 as often as every 5-10 min with less than 10(-5) breakthrough of Sr-82/85 over the 2- to 3-mo, useful life of the generator.