Coincidence Summing Research Articles

Gamma-gamma coincidence spectroscopy analysis with cascade summing corrections in close counting geometries

Gamma-gamma coincidence spectroscopy is a highly effective technique that becomes more challenging to implement when the distance between the detector and the sample is reduced. This difficulty arises due to cascade summing, which can complicate the acquisition of efficiency curves and introduce uncertainties, especially when using common sources like 152Eu. To address this issue, a method for correcting coincidence summing in close counting geometries is presented in this study. The approach utilizes MCNPX-Polimi and MCNP6.2, along with a 152Eu source calibration source for verification and validation. Additionally, the use of high purity germanium (HPGe) detectors coupled with Lynx digital signal analyzers (DSAs) for gamma-gamma coincidence is discussed. This work also outlines the generation, processing, and handling of time-stamped list (TLIST) mode data for dead-time corrections and coincidence gating.

A new methodology for efficiency calibration of uncharacterized Ge detectors by using characterized Ge detectors

The measurement of gamma emitters natural radionuclides is a very relevant subject in environmental radioactivity in many fields. For this, a valid calibration based on the full-energy peak efficiency (FEPE) is essential, which can be carried out for gamma-ray spectrometry using Monte Carlo codes (Geant4, LabSOCS, MNCP, etc.) without needing a source preparation and with much less time consuming compared to experimental methodologies. Therefore, this work aims to develop a methodology for the FEPE determination of uncharacterized Ge detectors by using characterized Ge detectors with known FEPE. For this, a general relationship was found between the simulated FEPE (εsim) for characterized Ge detectors by LabSOCS and the experimental one (εexp) for uncharacterized Ge detectors obtained for a problem sample (phosphate rock (PR) in our case), where εsim and εexp were factorized using three correction factors: geometrical, photon self-absorption and true coincidence summing (TCS) effects. Thus, the εsim/εexp ratio (FEPER) was obtained for two couples of Ge detectors (2 characterized and 1 uncharacterized detectors), using 4 different cylindrical geometries (C) and 2 Marinelli (M) ones, and varying the sample thickness (h), and gamma energy (Eγ) (from 46 keV (210Pb) to 1765 keV (214Bi)). The consistency obtained for the FEPERs for C geometries suggested a constant geometric factor, finding some deviations for M geometries only at Eγ < 150 keV. Given a pair of detectors, the behavior of the FEPERs was found to be very similar for all geometries, and h at each specific Eγ. Moreover, a constat relationship between the FEPERs of both detector couples versus Eγ was found. Then, the influence of the TCS effects was also studied varying the distance between the geometry bottom and detector window. The developed methodology was also validated confirming the consistency of the FEPERs, which were found to be independent on the chemical composition and apparent density of the sample, supporting our theoretical hypothesis. Consequently, given that the FEPER was found to be independent on the sample type and h, using our developed methodology it is possible to obtain εexp for any sample, geometry and h using the FEPER, previously obtained at each Eγ for the sample selected for methodology development (PR in our case), and the εsim that is obtained for the desired case.

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.

Investigation of 58Ni (n, p)58Co reaction cross-section with covariance analysis* *One of the authors (A.H.) sincerely acknowledges the Department of Science and Technology (DST), Government of India, for the INSPIRE Fellowship award (No. DST/INSPIRE Fellowship/2019/IF190924). G.S. and B.K.A. acknowledge partial support from the SERB, Department of Science and Technology, Government of India, with grants No. SIR/2022/000566 and CRG/2021/000101, respectively

The excitation function of the reaction was measured using the well-established neutron activation technique and γ-ray spectroscopy. Neutrons in the energy range of 1.7 to 2.7 were generated using the reaction. The neutron flux was measured using the standard monitor reaction. The results of the neutron spectrum averaged cross-section of reactions were compared with existing cross-section data available in the EXFOR data library as well as with various evaluated data libraries such as ENDF/B-VIII.0, JEFF-3.3, JENDL-4.0, and CENDL-3.2. Theoretical calculations were performed using the nuclear reaction code TALYS. Various nuclear level density (NLD) models were tested, and their results were compared with the present findings. Realistic NLDs were also obtained through the spectral distribution method (SDM). The cross-section results, along with the absolute errors, were obtained by investigating the uncertainty propagation and using the covariance technique. Corrections for γ-ray true coincidence summing, low-energy background neutrons, and γ-ray self attenuation were performed. The experimental cross-section obtained in the present study is consistent with previously published experimental data, evaluated libraries, and theoretical calculations carried out using the TALYS code.

Determination of detection efficiency on HPGe detector for point-like and volumetric samples based on Geant4 simulations

Production of medical isotopes has recently been developed at the Institute of Modern Physics (IMP) in China and many aqueous samples containing produced radioisotopes at wide range of volume were generated. Evaluation of radionuclidic purity and quantification of radioactivity for these samples are of significant importance, especially for medical purpose. High purity germanium (HPGe) detectors and gamma spectrometry are widely used in radionuclidic purity evaluation due to the high energy resolution. However, radionuclidic purity assessment and determination of radioactivity cannot be accurately conducted by using the HPGe detector installed at Laboratory of Nuclear Chemistry in IMP due to the lack of competent calibration standard sources. Therefore, an efficient method based on Geant4 simulation was proposed to calculate the detection efficiency curve for samples with different geometry, composition and source-to-detector distance. A set of efficiency data measured with a point-like standard source containing multi-radionuclides was used to determine the dimension of germanium crystal and thickness of dead layers. Afterwards, a point-like and 3 volumetric standard sources (5 ml, 10 ml and 20 ml, respectively) containing 152Eu were measured at 5–50 cm source-to-detector distances. The EFFTRAN code was applied to calculate the correction factor of coincidence summing for all measurements. The full energy peak (FEP) efficiencies with energy range of 121–1408 keV were calculated and compared with the values obtained from simulations. The simulated FEP efficiency values were in good agreement with experimental results, which illustrated a satisfactory precision of the Geant4 simulations. The feasibility of obtaining detection efficiency for point-like and volumetric samples by means of Geant4 simulations was confirmed. This method can be applied to address the efficiency problem caused by sample geometry and composition.

True coincidence summing corrections in HPGe detectors

The true coincidence effect is studied in two High Purity Germanium (HPGe) detectors for a variety of isotopes, source geometries and source to detector configurations, via computational tools based on Monte–Carlo simulations. Τhe upgraded patch of MCNP code MCNP–CP and the 2018 version of PENELOPE, which take into account the decay scheme of each cascade emitter, are used to calculate the Full Energy Peak Efficiency (FEPE) for the corresponding gamma-ray energies. The true coincidence correction (TCC) factor is calculated as the ratio of the FEPE derived for each nuclide taking into consideration the true coincidence effect, to the FEPE estimated without considering the phenomenon. In all cases, a satisfactory agreement is observed between the TCC factors calculated using MCNP–CP and PENELOPE 2018. Moreover, the results of the calculations are compared against experimentally derived efficiency values. The correction factors obtained using the TrueCoinc software are applied on experimentally determined FEPE curves, based on measurements performed using reference sources, and consequently the corrected data are compared against the simulations for the "non-coincidence" case. The results of this work contribute to the validation of the computational tools and codes used to study the true coincidence effect and determine the corresponding correction factors, providing useful data for gamma–spectrometry studies of cascade emitters.

Simultaneous correction of the coincidence summing and self-absorption for radioactivity measurement in solid samples by MCNP-CP code

Simultaneous correction of the coincidence summing and self-absorption for radioactivity measurement in solid samples by MCNP-CP code

Coincidence summing corrections using PENELOPE/PENNUC Monte Carlo code for volume sources of different densities

This work aims at providing a Monte-Carlo based methodology for calculating true coincidence correction (TCC) factors for volume sources of varying density. All simulations were carried out using the most recent version of Monte Carlo code PENELOPE. The main program PENMAIN was used for the calculation of full energy peak efficiencies. The subroutine PENNUC was utilized for the same calculation while taking summation effects into account. It was applied to Eu-152 and Cs-134 volume sources of 9 different densities, whilst the effect of the source's density on the TCC factor was investigated. There are differences between current results and the ones calculated by the TrueCoinc software. A relative bias up to 15% was observed, while the mean relative bias was 4.5%. The different approaches between the two codes could explain these deviations.

Development of concrete reference material for quality assurance/quality control of gamma radioactivity measurement for nuclear power plant decommissioning waste

The Korea Research Institute of Standards and Science has developed a new concrete reference material (RM) for measuring gamma-emitting radionuclides, such as 134Cs, 137Cs, 65Zn, 241Am, and 60Co, to improve and maintain the quality assurance and quality control of the radioactivity measurement in radioactive waste generated during the decommissioning of nuclear power plants. In this study, cement, SiO2, and bentonite, which are the main components of concrete, were mixed in an appropriate ratio with radionuclides. For certification and homogeneity assessment, 10 bottles were randomly selected, two sub-samples were collected from each bottle, and radionuclides were measured via HPGe gamma spectrometry. The results of the homogeneity tests using a one-way analysis of variance on 241Am, 134Cs, 137Cs, 65Zn, and 60Co in the concrete RM fulfilled the requirements of ISO Guide 35. Coincidence summing and self-absorption correction were performed on measurement results by introducing the Monte Carlo efficiency transfer code and Monte Carlo N-Particle transport code. The reference values for five radionuclides (60Co, 65Zn, 241Am, 134Cs, and 137Cs) in the RM were in the range of 15–40 Bq/kg, and the expanded uncertainty was within 10% (k = 2). To the best of our knowledge, this was the first study to develop concrete RM for measuring gamma-emitting radionuclides.

Corrections of HPGe detector efficiency curve due to true coincidence summing by program EFFTRAN and by Monte Carlo simulations

Standard reference sources, used for efficiency curve calibration of detector, often contain radionuclides with complex decay schemes (such as 60Co, 88Y, 152Eu …), introducing a potential problem in gamma-ray spectrometry, due to the appearance of coincidence summing of detected photons, in particular at a low source-detector distance. In this paper, a set of Monte Carlo simulations of an identical experimental setup were performed in order to obtain the efficiency curve of coaxial p-type HPGe detector for energy region (0–2) MeV, with the effect of true coincidence summing and without it. Obtained efficiency curves are compared with the experimental curve after applied EFFTRAN corrections. Fairly well agreement (between simulated and experimental curves with EFFTRAN corrections (with a relative deviation of 10%) proved the reliability of EFFTRAN corrections, as well as the possibility of Monte Carlo simulations for efficiency curve determination.

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.

Simplified efficiency calibration methods for semiconductor detectors used in criticality dosimetry

“Oversimplified” and “simplified” methods based on true coincidence summing effect used in uncomplicated determination of the photo-peak efficiency of the semiconductor High Purity Germanium (HPGe) detector system are suggested and verified. The methods and calibrated 60Co radioactive source may be used to commission any HPGe detector to use during potential criticality event. The determined accuracy of the semiconductor HPGe detector system using this method is a few percent (for the detector system used in this study it was ≃8% for “oversimplified” and ≃5% for “simplified” methods accordingly) reasonable, expected, and good enough to use for estimation of neutron dose from irradiated human blood in a potential criticality event. Moreover, if one can experimentally deduce the photo-peak efficiency for 60Co 1333 keV γ-ray line using the suggested methods, then with a few percent accuracy this efficiency could be also used for 1369 keV γ-ray line in the decay of 24Na isotope.

A Practical Procedure to Determine Natural Radionuclides in Solid Materials from Mining

There are many regulations related to the radiological control of NORMs (Naturally Occurring Radioactive Materials) in activities such as mining, industry, etc. Consequently, it is necessary to apply fast and accurate methods to measure the activity concentrations of long-lived natural radionuclides (e.g., 238U, 234,232,230,228Th, 228,226Ra, 210Pb, 210Po, and 40K) in samples characterized by a wide variety of compositions and densities, such as mining samples (wastes, minerals, and scales). Thus, it is relevant to calculate the radioactive index (RI), which summarizes for all radionuclides the ratio between the activity concentration and its respective threshold activity concentration as established by regulations, in order to classify a material as a NORM. To proceed with the determinations of these radionuclides, two spectrometric techniques based on both alpha-particle and gamma-ray detections should be employed. In the case of gamma-ray spectrometry, it is necessary to correct the full-energy peak efficiency (FEPE) obtained for the calibration sample, εc, due to self-attenuation and true coincidence summing (TCS) effects. The correction is especially significant at low gamma emission energies, that is, Eγ &lt; 150 keV, such as 46 keV (210Pb) and 63 keV (234Th). On the other hand, in samples which contain radionuclides that are in secular disequilibrium with others belonging to the same series (238U or 232Th series), like wastes or intermediate products, it is necessary to measure some pure-alpha emitters (232Th, 230Th, 210Po) by employing alpha-particle spectrometry. A practical and general validated procedure based on both alpha and gamma spectrometric techniques and using semiconductor detectors is presented in this study.

Calibration of cylindrical NaI(Tl) gamma-ray detector intended for truncated conical radioactive source

The computation of the solid angle and the detector efficiency is considering to be one of the most important factors during the measuring process for the radioactivity, especially the cylindrical γ-ray NaI(Tl) detectors nowadays have applications in several fields such as industry, hazardous for health, the gamma-ray radiation detectors grow to be the main essential instruments in radiation protection sector. In the present work, a generic numerical simulation method (NSM) for calculating the efficiency of the γ-ray spectrometry setup is established. The formulas are suitable for any type of source-to-detector shape and can be valuable to determine the full-energy peak and the total efficiencies and P/T ratio of cylindrical γ-ray NaI(Tl) detector setup concerning the truncated conical radioactive source. This methodology is based on estimate the path length of γ-ray radiation inside the detector active medium, inside the source itself, and the self-attenuation correction factors, which typically use to correct the sample attenuation of the original geometry source. The calculations can be completed in general by using extra reasonable and complicate analytical and numerical techniques than the standard models; especially the effective solid angle, and the detector efficiency have to be calculated in case of the truncated conical radioactive source studied condition. Moreover, the (NSM) can be used for the straight calculations of the γ-ray detector efficiency after the computation of improvement that need in the case of γ-γ coincidence summing (CS). The (NSM) confirmation of the development created by the efficiency transfer method has been achieved by comparing the results of the measuring truncated conical radioactive source with certified nuclide activities with the γ-ray NaI(Tl) detector, and a good agreement was obtained after corrections of (CS). The methodology can be unlimited to find the theoretical efficiencies and modifications equivalent to any geometry by essential sufficiently the physical selective considered situation.

Standardisation of 241Am activity for a key comparison

The JRC applied six measurement techniques to standardise the activity of an 241Am solution in the frame of the 2003 key comparison CCRI(II)-K2.Am-241. The methods used were alpha-particle counting at a defined small solid angle, high-efficiency particle and photon counting with a windowless 4π CsI(Tl) sandwich spectrometer, 4π alpha counting with a pressurised proportional counter, alpha-gamma coincidence counting and sum counting with a small pressurised proportional counter and a NaI(Tl) well detector, and 4π counting with a liquid scintillation counter. All results were consistent and an unusually low measurement uncertainty of 0.054% was achieved. An overview is presented of the outcome of the key comparison exercise, which demonstrates international equivalence.

Peak Efficiency of NaI Detector and Coincidence Summing Factor for Different Cylindrical Sources Using Geant4 Simulation.

Geant4 simulation is used to calculate peak efficiency and correct the effect of coincidence summing in detecting volumetric gamma-ray sources; this simulation was applied to a standard 152Eu source with different volumes as a test case. The source is a liquid cylindrical shape of various volumes. Peak efficiency was calculated using two tracks in the Geant4 simulation: single-energy track and "monoenergetic Track" without coincidence summing. Here, the energy of the source is known, and the track of radionuclides, including the coincidence summing, depends on the decay scheme of the radioactive source. The ratio between the peak efficiency of the two tracks gives us the correction factor (CF). The experimental method was used to calculate the peak efficiency and was amended by the correction factor computed with Geant4 tracks. The results showed a good agreement between experimental efficiency after correction and free-summing simulated efficiency. The comparison indicated that the present method is valid and useful for voluminous gamma-ray sources.

Efficiency calibration and coincidence summing correction for a NaI(Tl) spherical detector

Spherical NaI(Tl) detectors are used in gamma-ray spectrometry, where the gamma emissions come from the nuclei with energies in the range from a few keV up to 10 MeV. A spherical detector is aimed to give a good response to photons, which depends on their direction of travel concerning the detector center. Some distortions in the response of a gamma-ray detector with a different geometry can occur because of the non-uniform position of the source from the detector surface. The present work describes the calibration of a NaI(Tl) spherical detector using both an experimental technique and a numerical simulation method (NSM). The NSM is based on an efficiency transfer method (ETM, calculating the effective solid angle, the total efficiency, and the full-energy peak efficiency). Besides, there is a high probability for a source-to-detector distance less than 15 cm to have pulse coincidence summing (CS), which may occur when two successive photons of different energies from the same source are detected within a very short response time. Therefore, γ-γ ray CS factors are calculated numerically for a152Eu radioactive cylindrical source. The CS factors obtained are applied to correct the measured efficiency values for the radioactive volumetric source at different energies. The results show a good agreement between the NSM and the experimental values (after correction with the CS factors).

Efficiency calibration of high-purity germanium detector using Monte Carlo simulations including coincidence-summing corrections: volume source case

The gamma-ray spectrometry by the instrumentality of Ge detectors is used for the detection of low activity environmental samples of different geometry (soil samples, air filters with aerosols, milk powder, etc.). Such measurements require separate calibration of the detector. The high purity germanium (HPGe) gamma-ray spectrometer of GC2520 series was used for experiments. For the efficiency calibration, three cylindrical containers filled with different 60Co water solution levels were used, and the gamma-ray coincidence summing was modelled using MCNP6. The dimensions of the pure germanium crystal, provided by Canberra, were used for the simulations. The true coincidence summing takes place when two or more gamma quanta, which are emitted in a cascade from an excited nucleus, are detected within the resolving time of the detector. However, there is often a mismatch between the simulated and experimental efficiencies. The experimentally obtained and modelled spectra were compared: a good consistency of experimental and modelled results allows investigating the volume sources. During the simulation it was found that the factors affecting the accuracy of modelling are the thickness of the dead layer, crystal dimensions and the thickness of the Al detector cap. The analysis allows measuring the radionuclides activity concentration of samples placed in the containers with different filling heights having only standard shape calibration sources. The obtained accuracy is sufficient to fulfil criteria of 5–10% for such type of simulation to be applied for measurements of real samples in standard BURK-60 containers of various sample filling heights.

Study of coincidence summing effect using Monte Carlo simulation to improve large samples measurement for environmental applications

A new measurement model composed of High Purity Germanium detector HPGe including Lead Shield with a mixed source of Am-241, Cd-109, …in a soil matrix packed in a cylindrical container was developed by Monte Carlo N-Particles code MCNP5. The Monte Carlo models being largely used for efficiencies calculations in routine environmental radioactivity measurements of high volume soil samples.In order to validate the simulation, a multi-radionuclides soil matrix source in a cylindrical container is prepared. Comparing the simulated results to the experimental ones, allows us to correct coincidence summing in the soil standard and study its effect in activity measurements where a good agreement is found between corrected activities and results found by another geometry and show how uncorrected efficiency curves lead to erroneous activity calculation.Finally, the developed model is applied to study the distribution of natural and artificial radioactivity in a forest site in Bainem, northern of Algeria, these results are invested for radiological impact and soil erosion studies in the study Area.

Neutron radiative capture cross section for sodium with covariance analysis

The neutron radiative capture cross sections measurement has been carried out for the $$^{23}$$ Na nucleus in the neutron energy region from 0.6 to 3.2 MeV using the neutron activation technique followed by off-line $$\gamma $$ -ray spectrometry. The measurement was made relative to the $$^{115}$$ In(n,n $$\prime $$ $$\gamma $$ ) $$^{115}\hbox {In}^{m}$$ reference monitor reaction cross section. The neutrons were produced via the $$^{7}$$ Li(p,n) $$^{7}$$ Be reaction. Detailed uncertainty propagation has been performed using the covariance analysis, and the measured cross sections are being reported with their uncertainties, covariance, and correlation matrix. The necessary corrections have been made for the low background neutron energy contribution, $$\gamma $$ -ray true coincidence summing, and self-attenuation process. The obtained neutron spectrum averaged cross sections of $$^{23}$$ Na(n, $$\gamma $$ ) $$^{24}$$ Na are discussed and compared with the existing cross sections data retrieved from the EXFOR database. EMPIRE-3.2 and TALYS-1.9 calculations were performed in order to determine the radiative capture cross section in this energy region. The present results are also compared with the evaluated nuclear data from ENDF/B-VIII.0, TENDL-2019, IRDFF-1.05, JENDL-4.0, and JEFF-3.3. The obtained cross section results are in good agreement with existing experimental data, evaluated libraries, and reaction models for the highest energy points (2.11 and 3.13 MeV), while the lowest-energy point at 0.61 MeV underestimates them.