A NATIONAL SURVEY OF TRACEABILITY OF ACTIVITY METERS IN LITHUANIAN HOSPITALS.

The activity of prescribed radiopharmaceuticals, which are administered for diagnostic and therapeutic purposes, strongly depends on the accuracy of the measuring equipment used in nuclear medicine. This study presents the current status of uncertainty measurements of activity meters used in Lithuanian hospitals. During 2016-2021, the Ionizing Radiation Metrology Laboratory of the Center for Physical Sciences and Technology, which is the National Metrology Institute, performed a comparison of activity meters in the nuclear medicine departments of Lithuanian hospitals. Responses of 16 activity meters that were used in the daily hospital practice were compared with the reference standard. In total, around 150 measurements of activity of diagnostic, therapeutic and calibration radionuclides in different sources and geometries were fulfilled. The results of the maximum deviation were determined to be 8.9% for calibration sources, 18.3% for diagnostic radionuclides and 20.1% for radionuclides used for therapeutic purposes.

Occupational exposure in a PET/CT facility using two different automatic infusion systems.

PURPOSE: The aim of this study was to measure the occupational exposure using active personal dosimeters (APD) in the PET/CT department at different stages of the operation chain i.e. radiopharmaceutical arrival, activity preparation, dispensing, injection, patient positioning, discharge and compare the radiation exposure doses received using two automatic injection/infusion systems. This paper also reflects optimization processes that were performed to reduce occupational exposure. METHODS: Measured APD data were analysed for medical physicists, radiology technologists and administrative staff from 2014 till 2018. For dispensing and injecting 18F-FDG, the automatic infusion/injection system IRIDE (Comecer, Italy) or the automatic fractionator ALTHEA (Comecer, Italy) with wireless injection system WIS (Comecer, Italy) were used. Radiation exposure optimization methods were applied during the data collection period (installation of the transport port, patient management, APD alarm threshold and etc.). RESULTS: Radiology technologists who perform injection procedures, regardless of the automatic infusion system, received the highest radiation exposure dose. The average doses to the radiology technologists per one study were 1.72 ± 0.33 µSv and 1.16 ± 0.11 µSv with ALTHEA/WIS and IRIDE system, respectively. The average dose for accompanying the patient to the PET/CT scanner and scan procedure was 0.52 ± 0.07 µSv. For the medical physicists, the average dose was 0.29 ± 0.09 µSv. The measured dose for administrative staff was 0.30 ± 0.15 µSv. CONCLUSIONS: Occupational exposure can be effectively optimized by different means including staff monitoring with APD, implementation of radiation safety culture and the usage of automatic infusion systems.

Rapid analysis method for the determination of 14C specific activity in irradiated graphite.

14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1-100 µg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.

Uncertainty of determination of 158Tb in the RBMK nuclear reactor waste.

The activity of 158Tb was measured in waste samples from the Ignalina NPP Unit I RBMK-1500 reactor using gamma-ray spectrometry. The origin of 158Tb and the other observed gamma-ray emitters has been studied by using SCALE 6.1 modeling and comparing radionuclide ratios in the RBMK-1500 radioactive waste. The results of the calculation of the massic activity of gamma-ray emitters were used for interpretation of the total gamma-ray spectrum and the determination of 158Tb massic activity uncertainty in the waste of RBMK-1500.

Results of the CCRI(II)-S12.H-3 supplementary comparison: Comparison of methods for the calculation of the activity and standard uncertainty of a tritiated-water source measured using the LSC-TDCR method.

A comparison of calculations of the activity of a 3H2O liquid scintillation source using the same experimental data set collected at the LNE-LNHB with a triple-to-double coincidence ratio (TDCR) counter was completed. A total of 17 laboratories calculated the activity and standard uncertainty of the LS source using the files with experimental data provided by the LNE-LNHB. The results as well as relevant information on the computation techniques are presented and analysed in this paper. All results are compatible, even if there is a significant dispersion between the reported uncertainties. An output of this comparison is the estimation of the dispersion of TDCR measurement results when measurement conditions are well defined.

(3)H activity comparison between FTMC, VNIIM and LNE-LNHB.

An activity comparison of tritiated water was organized in 2013 between 3 laboratories: FTMC (Lithuania), LNE-LNHB (France) and VNIIM (Russia). The solution was prepared by LNHB and ampoules were sent to the others laboratories. This solution was standardized in terms of activity per unit mass by participant laboratories using the Triple to Double Coincidence Ratio (TDCR) method in liquid scintillation counting (LSC). The tritiated water solution is traceable to the solution prepared by LNHB for the CCRI(II)-K2.H-3 2009 (3)H international comparison.

Spatial patterns and ratios of ¹³7Cs, 9°Sr, and Pu isotopes in the top layer of undisturbed meadow soils as indicators for contamination origin.

Spatial distribution of activity concentrations of (137)Cs, (90)Sr, and (239,240)Pu in the top layer of undisturbed meadow soils was compared between two regional transects across Lithuania: one in the SW region, more affected by the Chernobyl radioactive fallout, and the other in the NE region. Radiochemical, γ-, α-, ß-, and mass spectrometric methods were used to determine the radionuclide activity. Our results validate that higher activity concentrations in the top soil layer were present in the SW region, despite the fact that sampling was performed after 22 years of the Chernobyl Nuclear Power Plant (NPP) accident. Using the activity concentration ratio (137)Cs/(239,240)Puglobal, the contribution of the Chernobyl NPP accident to the total radiocesium activity concentrations in these meadow soils was evaluated and found to be in the range of 6.5-59.1%. Meanwhile, the activity concentration ratio (238)Pu/(239,240)Pu showed that Chernobyl-derived Pu occurred at almost half of the sampling sites. The locations with maximal values of 47% of Chernobyl-derived Pu material were close to northeastern Poland, where deposition of most of non-volatile radioisotopes from the Chernobyl plume was determined.

Behaviour of (99)Tc in aqueous solutions in the presence of iron oxides and microorganisms.

A set of experiments was performed to determine the factors that influence TcO4(-) interaction with Fe-bearing minerals and to explore the effect of microbial activity on the behaviour of Tc(VII) in solution, in the presence of iron oxides under oxidizing medium. Gradual sorption of TcO4(-) (aq) onto wustite/magnetite was observed under alkaline conditions (pH 8-9). No pronounced effect of TcO4(-) (aq) interaction with hematite was observed in the investigating alkaline systems. At low pH values (2.7-4.5), TcO4(-) retention on hematite increases, suggesting that the process is dependent on pH. Sorption of (99)Tc (VII) onto hematite at pH 7.6-8.0 was achieved because of the presence of specific microorganisms.

Distribution of artificial gamma-ray emitting radionuclide activity concentration in the top soil in the vicinity of the Ignalina Nuclear Power Plant and other regions in Lithuania.

The impact of the operating Ignalina Nuclear Power Plant (INPP) on the contamination of top soil layer with artificial radionuclides has been studied. Results of the investigation of artificial gamma-ray emitting radionuclide distribution in soil in the vicinity of the INPP and distant regions in Lithuania in 1996-2008 (INPP operational period) show that nowadays (137)Cs remains the most important artificial gamma-ray emitting radionuclide in the upper soil layer. Mean (137)Cs activity concentrations in the top soil layer in the vicinity of the INPP were found to be significantly lower compared to those in remote regions of Varena and Plunge (~300 km from INPP). In 1996 and 1998 mean (137)Cs activity concentrations were in the range of 28-45 Bq/kg in the nearest vicinity to the INPP, 103 Bq/kg in Varena and 340 Bq/kg in Plunge region. (137)Cs activity concentrations were 5-20 times lower in meadow soil (4-14 Bq/kg) compared to swamp and forest soil. (60)Co, the INPP origin radionuclide, was detected in samples only in 1996 and 2000, and the activity concentration of (60)Co was found to be in the range from 0.4 to 7.0 Bq/kg at the sampling ground nearest to the INPP. Average annual activity concentrations of the INPP origin (137)Cs and (60)Co in the air and depositions in the INPP region were modeled using Pasquill-Gifford equations. The modeling results of (137)Cs and (60)Co depositional load in the INPP vicinity agree with the experimentally obtained values. Our results provide the evidence that the operation of INPP did not cause any significant contamination in soil surface.

Calibration of stack monitors for measurement of noble gases in nuclear facilities.

In nuclear facilities stack monitors are used for the measurement of the volumetric activity of noble gases. Spectrometric measurement is needed because the content of stack effluents is always a mixture of radionuclides. In some nuclear power plants new types of monitors were installed based on HPGe detectors. For efficiency calibration a standard with the radionuclide Xe-127 was developed and calibration curve constructed in the energy range 81 keV-1293 keV. Experiental efficiencies were checked using an MC model.

Laboratories new to the ICRM.

The Scientific Committee of the ICRM decided, for the 2011 Conference, to present laboratories that are at a key developmental stage in establishing, expanding or applying radionuclide metrology capabilities. The expansion of radionuclide metrology capabilities is crucial to meet evolving and emerging needs in health care, environmental monitoring, and nuclear energy. Five laboratories (from Greece, Lithuania, Indonesia, Norway and Turkey) agreed to participate. Each laboratory is briefly introduced, and examples of their capabilities and standardization activities are discussed.

Total efficiency of GE detectors--dead layer signal effect.

Two types of detectors (GC 4018 and BE 5030) were compared regarding the signal from the outer dead layer of the detector. Gamma-spectra of Am-241 were acquired with various delays against the starting signal from an alpha LS probe. Coincidence spectra from both detectors were different. The dead-layer signal does not increase the full-energy peak efficiency but it increases the total efficiency, as it was demonstrated by the reconstruction of Am-241 spectra for both detectors.