Marinelli Beaker Research Articles

An operational approach for accurate 177Lu and 177mLu activity quantifications to comply with the environmental release criteria: the role of GEANT4 for efficiency curve and True Coincidence Summing effect estimation

The increasing use of 177Lu-labelled agents for targeted radionuclide cancer therapy highlights the radiation protection challenges in the management of radioactive waste due to 177mLu impurities. To ensure safe disposal, compliance with clearance criteria must be verified using calibrated systems, such as Hyper Pure Germanium (HPGe) spectrometers. This work aims to validate a customized GEANT4 model of our HPGe system in Marinelli beaker geometry to propose an operational approach properly quantifying 177Lu and 177mLu in waste samples. The system's efficiency curve was calculated by modelling gamma sources in the energy range of interest and validated by spectra measurements of 177Lu and 99mTc sources. Correction factors accounting for true coincidence summing (TCS) effect were simulated for 177Lu, 177mLu and 99mTc and they were applied to the spectrum measurement of a waste sample model with known activities of 177Lu and 177mLu. Thus, an operational approach for activities quantification was tested comparing the results with the nominal values. No significant differences were observed between simulated and measured efficiency values. TCS correction factors are significant only for 177mLu (1.6 at 112.95 keV and 204.11 keV). Eventually, the proposed framework to quantify 177Lu in a waste sample allowed to estimate the 177Lu and 177mLu component activities within a maximum 16% uncertainty. Results show that the HPGe model could be a powerful resource for a wide range of applications in daily clinical routine and it could be used to build a simple quality assurance program to monitor the detector response constancy in time.

A semi–empirical method for efficiency calibration of an HPGe detector against different sample densities

In this work, a semi-empirical equation in terms of γ-energy, and sample density is derived, proposed, benchmarked, and applied for the peak efficiency calibration of an HPGe detector with respect to an axial source-to-detector configuration. The samples are in the form of cone-shaped Marinelli beakers of different densities in the range 0.7–1.6 g/cm3. The method employs the experimental measurements with the ANGLE-3 software calculations using the efficiency transfer method. The peak efficiency curve of an HPGe detector is calculated using the experimental measurements of point-like sources (133Ba, 137Cs, and 60Co). The ANGLE-3 software is then used to calculate the peak efficiency curves for samples with different densities in the γ-energy range 81–1332 keV. The peak efficiency curves are then fitted to get the energy coefficient; in addition, a linear relationship is then constructed between the energy coefficients and sample densities to get the density coefficients, and the derived equation as well.The derived equations are benchmarked using the peak efficiency curves by ANGLE-3 software in comparison with that the equation results. The results are found to be in agreement with an average relative error of about 1.5%. In addition, the derived equations are applied to estimate the activity concentration of radionuclides present in 5 cone-shaped samples with different densities using experimental measurements. The activity results are found to be in agreement with the certified values with an average relative error of about 2%. The limitation of the proposed equations is also discussed with respect to different material densities and different chemical compositions and correction factors for material composition self-attenuation for various materials are also presented.

X-ray radiography of HPGe detectors for Monte Carlo simulations

The results of the development of a radiographic installation that allows to control the size of the internal elements of HPGe detectors' cryostats for simulation by Monte Carlo method, are presented. Calibration and assessment of image quality value of the developed radiographic installation were carried out in accordance with ISO Standards 19232–1 and 19232–5. Simulation of the HPGe detector's registration efficiency in the range of 50 keV-3 MeV with the MCC-MT program showed that the discrepancy between the calculated and experimental results when adjusting the detector model using radiological measurement data does not exceed 3% in geometry for point sources and 5% for the Eu-152 source in the Marinelli beaker geometry.

A New Approach for NaI(Tl) Detector Efficiency Calibration Using 41Ar Radionuclide for Air Exhaust Systems

Argon-41 is an essential gaseous radionuclide that must be monitored in gaseous effluents from nuclear facilities. Therefore, a precise evaluation of 41Ar activity is highly desired. Gamma spectroscopy with a NaI(Tl) scintillation detector coupled with a multichannel analyzer (MCA) is one of the widely used techniques for the identification and activity measurements of radioisotopes. However, the efficiency calibration of these kinds of monitoring systems highly depends on the source-detector geometry, and a large amount of uncertainty may complicate the calibration. This paper presents the evaluation of the full peak efficiency of a 2 × 2-in. NaI(Tl) scintillation detector coupled with a stable MCA for a 41Ar source with 1293.5 keV energy in two different source-detector geometries, duct and Marinelli beaker, using the FLUKA code. A new experimental technique is considered to produce 41Ar in a controlled geometry, like a Marinelli beaker, through neutron irradiation of natural argon inside a cyclotron bunker. The simulation data were compared with the experimental results for Marinelli beaker geometry, and the ratio was evaluated as 0.99 ± 0.07. The ratio was considered a scaling factor for the final efficiency calibration of duct geometry.

Evaluating coincidence summing factor using marinelli beaker on coaxial HPGe detector by Monte Carlo simulation and calculating

This investigation aims to compare the full energy peak efficiencies in the energy range of 46-1836 keV on a type-p coaxial HPGe and estimate the coincidence summing factor for the case of Marinelli Beaker samples used by two general Monte-Carlo simulation software MCNP and PENELOPE. The radioactive nuclides used in determining the coincidence summing factor include 22Na, 60Co, 88Y, 133Ba, 134Cs, 154Eu, and 208Tl, which are prepared in HCl 2M solution and contained in a Marinelli beaker with the source’s volume of 3000 ml. The results demonstrate there is a good agreement between the two simulation software with an average discrepancy of 1.3%. On the other hand, the simulation coincidence summing factor values are also compared with the results from the calculating software ETNA with an average discrepancy of approximately 3.1%.

Comparison of Experimental and Monte Carlo Efficiencies of 0.5g/cc Epoxy Matrix Marinelli Source with Multiple Radioactive Nuclides

Using the gamma spectroscopy system, it can be determined whether environmental samples or standard radioactive sources are radioactive, and from which elements their radioactivity originates. The purpose of Monte Carlo (MC) simulation is to model a real-life system with its inputs and evaluate the outputs with real results. This study calculates the experimental efficiency of a p-type HPGe detector using a 0.5 g/cc Epoxy Matrix Marinelli beaker and compares these results with GESPECOR and PHITS MC Simulation programs. Thus, the thickness of the dead layer, which thickens over time and affects the detector efficiency, was determined from the most compatible result of the MC calculations made repeatedly at various alternative thicknesses to the experimental results. For 1.5 mm dead layer thickness, less than 2 % error was found between the test and MC results, especially at energies above 165 keV. As a result, it was determined that the dead layer thickness of the detector reached 1.5 mm with an increase of 114 % after its production. The current value of the dead layer thickness of each detector should be checked, as the efficiency affects the determination of the activity.

Dead layer estimation of an HPGe detector using MCNP6 and Geant4

In this study, Monte Carlo simulations were used to calculate the full-energy peak efficiency of a p-type coaxial high-purity germanium (HPGe) detector. The HPGe detector was modeled using MCNP6 and Geant4, and the thickness of the dead layer of germanium crystals was estimated for an accurate simulation. The dead layer was divided into front and side components, where a point source and a Marinelli beaker source were used to estimate each dead layer thickness. The model was validated by comparing the simulated as well as experimental results for the standard sources of cylindrical and Marinelli beakers. The Geant4 results and experimental results matched up to 4% in the 59.54–1836.05 keV energy range, while MCNP6 matched up to 6% when adjusted for coincidence summing effects. HPGe detector modeled in Monte Carlo simulations can be utilized for experimental validation and experimental setup prior to using actual HPGe detectors.

COMPARISON BTWEEN THE RELATIONS OF HpGe DETECTOR EFFICIENCY CURVE AND BACKGROUND “SPECTRUM SHAPE”

The full energy peak efficiency (ɛ) relation of HpGe detector is determined and compared with the "spectrum shape" relation of natural background (BG) for the (121.8-1408) keV energy range. A one liter Marinelli beaker soil standard containing the Eu-152 material is used in calibration and measurement. A MATLAB fitting program were used obtain the efficiency relation from the experimental values, and to obtain the BG "spectrum shape" relation from the count rates at the corresponding energy values used. The ɛ / BG factor values were 1.154 at the 121.8keV and increase with energy till reaching a maximum value of 4.358 at about 778keV energy and then decrease. These values may be used to extract efficiency from the BG values at the specified measurement conditions. The obtained results may participate in extending our understanding of the standardization concept in radiation measurements.

Determination of the full energy peak efficiency of HPGe detector for varying soil compositions and Marinelli beaker dimensions using Artificial Neural Networks

Natural radioactivity measurement is studying the radiation emitted naturally from any nature's components such as air, soil, water, etc. This field of study plays an important role in public health and safety management. Scientists and engineers can calculate the radiation doses received by inhabitants of a certain area by measuring the natural radioactivity in the same area. One of the main issues facing the standard measurement procedure is determining the appropriate volume of the sample to be taken from the area under investigation. The volume of the investigated sample affects the detection efficiency of the HPGe detector due to the self-shielding effect. In this work we show how using artificial neural networks can help in choosing the proper dimensions of the Marinelli beaker to be used in the detection process. We used the various compositions of soil as a case study. Results showed that artificial neural networks were capable of finding the relation between soil composition, Marinelli beaker dimensions, and gamma energy as inputs and the HPGe detection efficiency as output. The developed model's regression was 99% compared to the detection efficiency calculated using MCNP. This will help to have a better prediction of the detection efficiency of natural radioactivity measurements.

MODELING AND OPTIMIZATION OF HPGE DETECTOR GC0518 USING MCNP5 CODE.

For precise modeling of a high purity germanium (HPGe) detector, detailed knowledge of its geometry is necessary. Unfortunately, some of the detector geometric parameters are often not available. This study suggests simple procedures for determining geometry without the need for radiography. The study considers the HPGe detector GC0518 manufactured by Mirion. The detector datasheet is available, but the uncertainties do not allow the detector to be modeled without additional corrections. The detector was modeled using MCNP5 code and the results were compared with calibration point sources and Marinelli beakers measurements. We also performed measurements to obtain information on the germanium dead layer, crystal holder shape, inner hole shape and radii of the crystal edges. Then the model was optimized according to the measurements. Finally, the full energy peak efficiencies were evaluated. The relative differences between the measured and calculated efficiencies are less than 4.5% for point sources and 5.4% for Marinelli beakers.

Coincidence Summing Factor Calculation for Volumetric γ-ray Sources Using Geant4 Simulation

Geant4 simulation was applied to correct the coincidence summing (CS) effect in detecting a volumetric γ-ray sources, and this technique was applied to a152Eu standard sources. The radioactive sources were a liquid cylindrical, rectangular, and Marinelli beaker shapes of different volume for each one. Radionuclide track (RT) including coincidence summing and monoenergetic track without coincidence summing. The results obtained from two approaches compared with the experimental data and the modified KORSUM code for cylindrical γ-ray source. The comparison showed that the adopted method in this investigation is useful for coincidence summing corrections for a voluminous γ-ray sources.Moreover, this technique requires far less computation time than the techniques that depend on the calculation of total efficiency.

Эзэлхүүнт дээжийг судлах гамма-спектрометрийн үнэмлэхүй арга

The Absolute Method for determination of gamma-emitting radionuclides in environmental samples using gamma spectrometry was described. Density correction coefficients of Gamma:Ray Detection efficiency taking into account absorption of gamma-quants in environmental samples were calculated by the this method (standard Marinelli beaker geometry).

Development of Neural Network Model With Explainable AI for Measuring Uranium Enrichment

In this work, we have developed a neural network (NN) model that can analyze enrichment from depleted (0.2%) to low enriched uranium (4.5%) when UO <sub xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink">2</sub> waste with very low radioactivity was contained in a 1-L Marinelli beaker, even when the measurement time is as short as 10 s using a low-resolution detector. The average count rate was about 3800 cps. Measurement of uranium enrichment is necessary for quantitative analysis of uranium radioactivity for disposal of uranium waste. Previously studied uranium enrichment methods (infinite thickness (IT) method, peak ratio (PR) method, and relative-efficiency (RE) method) are difficult to use for field measurement due to many limitations of the algorithms. Among existing methods, the RE method is accurate but requires a long measurement time; there is also a limitation in which a high-resolution detector is essential. In this work, we proposed a model to predict uranium enrichment using a low-resolution detector and an artificial NN model. Furthermore, we validated the results of the NN models using an explainable AI algorithm and principal component analysis (PCA). When the measurement time was less than 60 s, the existing method failed to analyze uranium enrichment, but the proposed model can predict enrichment of uranium within 5% of relative error when 5 g of uranium powder was mixed with various waste (ash, soil, and concrete).

Determination of 226Ra in TENORM Sample Considering Radon Leakage Correction

Background: Phosphogypsum is material produced as a byproduct in fertilizer industry and is generally used for building materials. This material may contain enhanced radium-226 (226Ra) activity concentration compared to its natural concentration that may lead to indoor radon accumulation. Therefore, an accurate measurement method is proposed in this study to determine 226Ra activity concentration in phosphogypsum sample, considering the potential radon leakage from the sample container.Materials and Methods: The International Atomic Energy Agency (IAEA) phosphogypsum reference material was used as a sample in this study. High-purity germanium (HPGe) gamma spectrometry was used to measure the activity concentration of the 226Ra decay products, i.e., 214Bi and 214Pb. Marinelli beakers sealed with three different sealing methods were used as sample containers. Due to the potential leakage of radon from the Marinelli beaker (MB), correction to the activity concentration resulted in gamma spectrometry is needed. Therefore, the leaked fraction of radon escaped from the sample container was calculated and added to the gamma spectrometry measured values.Results and Discussion: Total activity concentration of 226Ra was determined by summing up the activity concentration from gamma spectrometry measurement and calculated concentration from radon leakage correction method. The results obtained from 214Bi peak were 723.4 ± 4.0 Bq· kg-1 in MB1 and 719.2 ± 3.5 Bq· kg-1 in MB2 that showed about 5% discrepancy compared to the certified activity. Besides, results obtained from 214Pb peak were 741.9 ± 3.6 Bq· kg-1 in MB1 and 740.1 ± 3.4 Bq· kg-1 in MB2 that showed about 2% difference compared to the certified activity measurement of 226Ra concentration activity.Conclusion: The results show that radon leakage correction was calculated with insignificant discrepancy to the certified values and provided improvement to the gamma spectrometry. Therefore, measuring 226Ra activity concentration in TENORM (technologically enhanced naturally occurring radioactive material) sample using radon leakage correction can be concluded as a convenient and accurate method that can be easily conducted with simple calculation.

Content of cesium-137 in the soil cover of Central and Southern Tajikistan

The content of cesium-137 in the soil cover of the central and southern parts of Tajikistan was studied. The study area of the regions of the republic is about 45,000 km2. 92 soil samples from the Central and Southern parts of the republic were collected from cultivated and uncultivated fields to a depth of 25 cm from the ground surface. Each sample at the sampling point was cleared of stones, roots and other inclusions. The sample was dried to an air-dry state, crushed and sieved through a sieve with a hole diameter of 2 mm. When measuring the samples, we used a standard 1 L Marinelli beaker. Measurements of the specific activity of cesium-137 in soil samples were carried out on a gamma spectrometer based on highly pure germanium. The maximum content of cesium-137 was observed in soil sample No. 1 of the Kamarob area of the Rasht district, which specific activity in the sample is 148 Bq / kg. The minimum specific activity of cesium-137 is observed in soil samples from the Tajikabad district. The dependence of the concentration of cesium-137 on the topography in the region under study has been established. In some of the investigated samples, the cesium-137 isotope was not detected (sample no. 2 from the Ayvaj area of the Shahritus district and sample no. 4 from the Farkhor district). The dynamics of the distribution of cesium-137 on the soils of uncultivated fields and soils of cultivated fields of the Khatlon region of Tajikistan has been studied. It was found that the content of the isotope of cesium-137 in the treated fields differs from the untreated ones. It has been determined that in the soils of cultivated lands, the content of cesium-137 is concentrated at a depth of 20-25 cm. At the foot of the mountains and clay soils, a relatively high content of the isotope of cesium-137 is observed in comparison with plains and sandy soils.

Radioactivity measurement of 85Kr by using anti-cosmic HPGe γ spectrometer

In order to improve the detection sensitivity of 85Kr, an anti-cosmic HPGe γ spectrometer was established and a specific Marinelli beaker was designed for 85Kr measurement. Comparing to the non-anticoincidence γ spectrum, the integral background counts rate range from 20 to 2400 keV was reduced by a factor of 5.17 using anticoincidence shielding. The minimum detectable activity of 85Kr was 10.5 Bq within 24 h measurement in standard mode and that of 3.99 Bq in anticoincidence mode. The activity concentration of atmospheric 85Kr was preliminarily measured to be an average of 1.30 Bq/m3 in Beijing and corresponding minimum detectable concentration was 0.2 Bq/m3.

Activity Concentrations Of Environmental Samples Collected In Samarkand Region Of Uzbekistan

The results of the investigations of the activity concentration of natural radionuclides in certain building materials and foodstuff samples obtained from the Nurabad district of the Samarkand region of Uzbekistan are presented. The gamma radiation spectra of the samples were measured in Marinelli beaker geometry on γ- spectrometer with a Ge(Li) detector and energy resolution 6% on the 1332 keV line of 60Co. A personal computer and standard computer programs were used to accumulate and process the spectra. The characteristic density of the samples ranged from 140 to 1810 g/l. The activity concentration of 40K was observed to be comparatively higher than that of both 226Ra and 232Th in all the studied samples. The results reveal that the low activity of 137Cs in the samples is attributed due to its half-life decay period.

Surface Contamination and Dose Rate Verification of Fertilizers common in Iraqi Plantations using RadEye B20 Detector

Twenty three samples of granular chemical fertilizers and organic fertilizers commonly utilized in Iraqi ranches were collected. The samples were prepaid and stored in a Marinelli beaker to measure; dose rate, general count rate and surface contamination of the samples using the RadEye B20 detector, firstly with shield, secondly without the shield to estimate the effect of shielding on the measurements. The results showed that using shield made a significant decrease in the radiation measurements reached about 25%. However the mean value of surface contamination, dose rate and general count rate with shield were 0.54Bq/cm2, 0.65µsv/h, and 0.28Cps respectively, and without shield being 0.34Bq/cm2, 1.33µsv/h, and 1.52Cps respectively.

Prototype Mn-solution bath for low activity primary standard neutron dosimetry

In this work, a prototype Mn-bath was designed and constructed in our laboratory. It is composed of a spherical 304 stainless steel vessel with a diameter of 34 cm. A hydraulic system for circulating the 55Mn solution between the bath and the detection system. The detection system is composed of a Marinelli beaker attached to a 3 × 3″ NaI(TI) crystal detector connected to the Nomad analyzer. A 252Cf source with a flux of order 103 (n cm−2 S−1) was used. The output dose was measured by a standard NM2 neutron monitor. The output flux was calculated from the flux-dose relationship. It was found that the flux delivered by the used 252Cf was 170.56 ± 6.41 n cm−2 S−1as calculated from the dose measured by NM2. Different concentrations of 55Mn solution with the molarity of 0.89,1.18 and 1.48 (mol/l) extracted a flux of 194.48, 186.46, and 169.79 n cm−2 S−1, respectively. Data of neutron flux obtained from concentrations of 1.18 and 1.48 mol/l was in a good agreement with the flux measured by the standard neutron monitor and the guaranteed value of the used source. Uncertainty of flux measured by Mn bath was enhanced with increased concentration.

Design, construction and characterisation of a portable gamma-ray spectrometer for low-level natural occurring radioactive material ex-situ measurement

The activity concentrations in naturally occurring radioactive material (NORM) samples are conventionally measured using a gamma-ray spectrometer with a single detector (mostly HPGe or NaI:Tl) enclosed in a lead shield. In this work, a passive water shield to reduce background radiation reaching the detectors was designed using GEANT4-toolkit Monte Carlo simulations and then constructed. This measurement system is portable and cost-effective for ex-situ measurements. IAEA-375 soil and beach sand each placed in Marinelli beakers were measured using two LaBr3:Ce detectors in singles and coin-cidence modes. A novel method of background reduction by using photon time-of-flight was employed together with the measurement of the two photons energies. These samples were also measured in singles mode using a NaI:Tl detector inside the constructed water shield and HPGe detector shielded with lead to compare and validate the results of LaBr3:Ce detectors measurement. Both the simulated and measured results show that the water shield (500 mm thick) attenuates 2614.5 keV gamma rays by 90 %. The minimum detectable activity (MDA) measured for 24 h using the NaI:Tl detector without shield and inside the water shield are; 546 Bq kg−1 and 146 Bq kg−1 at 1460.8 keV (40K), 194 Bq kg−1 and 15 Bq kg−1 at 1764.5 keV (238U series), and 131 Bq kg−1 and 15 Bq kg−1 at 911.2 keV (232Th series), respectively. The measured activity concentrations of 238U and 232Th series and 40K radionuclides inside IAEA-375 soil agree with certified values to within measurement uncertainties. The measured activity of 138La in the LaBr3:Ce detector crystal is 263.8 ± 26.8 Bq kg−1. The internal activity of the LaBr3:Ce detector increases the MDA at 1460.8 keV and 2614.5 keV, which limits the measurement of primordial radionuclides with low activity concentration in singles mode. The activity concentrations of 238U and 232Th series radionuclides in beach sand were measured using NaI:Tl and LaBr3:Ce detectors. The results are consistent with those from HPGe measurement to within uncertainties.