ABSTRACT
BACKGROUND: Radiopharmaceuticals are considered as regular medicinal products and therefore the same regulations as for non-radioactive medicinal products apply. However, specific aspects should be considered due to the radiochemical properties. Radiopharmaceutical dedicated monographs are developed in the European Pharmacopoeia to address this. Currently, different quality control methods for non-registered radiopharmaceuticals are utilized, often focusing on radio-TLC only, which has its limitations. When the radiochemical yield (RCY) is measured by radio-TLC analysis, degradation products caused by radiolysis are frequently not detected. In contrast, HPLC analysis defines the radiochemical purity (RCP), allowing for detection of peak formation related to radiolysis. During the introduction and optimization phase of therapeutic radiopharmaceuticals, significant percentages of impurities, like radiolysed construct formation, may have consequential impact on patient treatment. Since more hospitals and institutes are offering radiopharmaceutical therapies, such as [177Lu]Lu-PSMA with an in-house production, the demand for adequate quality control is increasing. Here we show the optimization and implementation of a therapeutic radiopharmaceutical, including the comparison of ITLC and HPLC quality control. RESULTS: Downscaled conditions (74 MBq/µg) were in concordance to clinical conditions (18 GBq/250 µg, 5 mL syringe/100 mL flacon); all results were consistent with an > 98% RCY (radio-TLC) and stability of > 95% RCP (HPLC). Radio-TLC did not identify radiolysis peaks, while clear identification was performed by HPLC analysis. Decreasing the RCP with 50%, reduced the cell-binding capacity with 27%. CONCLUSION: This research underlines the importance of the radiolabeling and optimization including clinical implementation and clarifies the need for cross-validation of the RCY and RCP for quality control measurements. Only HPLC analysis is suitable for identification of radiolysis. Here we have proven that radiolysed [177Lu]Lu-PSMA has less binding affinity and thus likely will influence treatment efficacy. HPLC analysis is therefore essential to include in at least the validation phase of radiopharmaceutical implementation to ensure clinical treatment quality.
ABSTRACT
Objective:To explore the assessment methodology for internal dose to workers exposed to 131I radionuclide. Methods:Workers were chosen in a 131I radiopharmaceutial manufacturer and a nuclear medicine department in a hospital using 131I to treat hyperthyroidism and thyroid cancer. A portable high purity germanium (HPGe) gamma spectrometer was used to measure the content of 131I in the thyroid for 4 consecutive times in a period of 7 d. The internal dose was estimated combining with the work rotation mode for workers dealing with 131I. Results:When the monitoring month was used as a typical month to estimate the internal dose, the annual committed effective dose was 0.09-1.93 mSv for the production staff engaged in the repackaging of 131I radiopharmaceuticals in the surveyed enterprise, and 0.06-0.58 mSv for the nuclear medicine staff in the surveyed hospital. After adjusting the monitoring result of the current monitoring period based on the rotation mode, the annual committed effective dose was estimated to be 0.06-1.22 mSv for radiopharmaceutical production workers and 0.03-0.15 mSv for nuclear medicine workers, respectively. Conclusions:In the assessment of internal dose to radiation workers exposed to 131I, using a single time measurement result to estimate the annual dose would lead to a larger error. In the case of continuous monitoring, the result of subsequent monitoring periods should be corrected according to the result of previous monitoring periods. In order to accurately estimate the internal dose of workers exposed to 131I, it is necessary to take full account of the 131I exposure pattern, time and frequency and the internal contamination route. For workers who may be exposed to 131I with potential internal dose greater than 1 mSv/year, a 14 day-routine monitoring period was appropriate.
ABSTRACT
Technetium-99m (99mTc) is the most used radionuclide worldwide in nuclear medicine for diagnostic imaging procedures. 99mTc is typically extracted from portable generators containing 99Mo, which is produced normally in nuclear reactors as a fission product of highly enriched Uranium material. Due to unexpected outages or planned and unplanned reactor shutdown, significant 99mTc shortages appeared as a problem since 2008 The alternative cyclotron-based approach through the 100Mo(p,2n)99mTc reaction is considered one of the most promising routes for direct 99mTc production in order to mitigate potential 99Mo shortages. The design and manufacturing of appropriate cyclotron targets for the production of significant amounts of a radiopharmaceutical for medical use is a technological challenge. In this work, a novel solid target preparation method was developed, including sputter deposition of a dense, adherent, and non-oxidized Mo target material onto a complex backing plate. The latter included either chemically resistant sapphire or synthetic diamond brazed in vacuum conditions to copper. The target thermo-mechanical stability tests were performed under 15.6 MeV proton energy and different beam intensities, up to the maximum provided by the available GE Healthcare (Chicago, IL, USA) PET trace medical cyclotron. The targets resisted proton beam currents up to 60 µA (corresponding to a heat power density of about 1 kW/cm²) without damage or Mo deposited layer delamination. The chemical stability of the proposed backing materials was proven by gamma-spectroscopy analysis of the solution obtained after the standard dissolution procedure of irradiated targets in H2O2.
Subject(s)
Cyclotrons , Technetium/chemistry , Cyclotrons/instrumentation , Radiopharmaceuticals , Spectrometry, GammaABSTRACT
A new regulatory statement was issued concerning the eye lens radiation protection of persons in planned exposures. A debate was raised on the adequacy of the dosimetric quantity and on its method of measurement. The aim of this work was to establish the individual monitoring procedure with the EYE-D™ holder and a MCP-N LiF:Mg,Cu,P thermoluminescent chip detector for measuring the personal dose equivalent Hp(3) in workers of a Positron Emission Tomography Radiopharmaceutical Production Facility.
