HPGe Detector Research Articles

Characterisation of Two P-Type Hpge Detectors by Measurements and Monte Carlo Simulations

Characterisation of Two P-Type Hpge Detectors by Measurements and Monte Carlo Simulations

New limits on double-beta decay of ^{190}Pt and ^{198}Pt

A search for double-beta decay of ^{190}Pt and ^{198}Pt with emission of gamma -ray quanta was realized at the HADES underground laboratory with a 148 g platinum sample measured by two ultralow-background HPGe detectors over 8946 h. The isotopic composition of the platinum sample has been measured with high precision using inductively coupled plasma mass spectrometry. New lower limits for the half-lives of ^{190}Pt relative to different channels and modes of the decays were set on the level of lim T_{1/2}sim 10^{14}–10^{16} year. A possible exact resonant 0nu KN transition to the 1,2 1326.9 keV level of ^{190}Os is limited for the first time as T_{1/2} ge 2.5 times 10^{16} year. A new lower limit on the double-beta decay of ^{198}Pt to the first excited level of ^{198}Hg was set as T_{1/2} ge 3.2times 10^{19} year, one order of magnitude higher than the limit obtained in the previous experiment.

AGE-DEPENDENT DOSE DUE TO INTAKE OF 40K, 228RA, 226RA IN SPRING WATER IN PETRA DISTRICT, JORDAN.

A HpGe detector was used to measure radioactivity concentrations of 40K, 228Ra and 226Ra in spring water from villages in Petra district in southern Jordan. The concentrations of 40K, 228Ra, 226Ra were 0.64-2.72Bq/l, 0.05-0.08Bq/l and 0.15-0.22Bq/l, respectively. These values were compared to the concentrations of the corresponding radionuclides in studies from Jordan and near countries. The annual effective doses (${D}_{\mathrm{eff}}$) due to the intake of 40K, 228Ra and 226Ra, for the different age groups were calculated. The highest ${D}_{\mathrm{eff}}$ values due to the intake of 226Ra and 228Ra were found in the infants age group, whereas the lowest were found in the adults age group in every site. The annual effective doses in this study were compared to the committed annual effective doses from ingestion in UNSCEAR. The annual effective doses in this study were much higher than the committed values in UNSCEAR. The life time risk for radiation-induced cancer for the whole population was calculated for every sample and it revealed no extra risk over the one recommended by WHO.

Enhancement of Ceramics Based Red-Clay by Bulk and Nano Metal Oxides for Photon Shielding Features.

We prepared red clays by introducing different percentages of PbO, Bi2O3, and CdO. In order to understand how the introduction of these oxides into red clay influences its attenuation ability, the mass attenuation coefficient of the clays was experimentally measured in a lab using an HPGe detector. The theoretical shielding capability of the material present was obtained using XCOM to verify the accuracy of the experimental results. We found that the experimental and theoretical values agree to a very high degree of precision. The effective atomic number (Zeff) of pure red clay, and red clay with the three metal oxides was determined. The pure red clay had the lowest Zeff of the tested samples, which means that introducing any of these three oxides into the clay will greatly enhance its Zeff, and consequently its attenuation capability. Additionally, the Zeff for red clay with 10 wt% CdO is lower than the Zeff of red clay with 10 wt% Bi2O3 and PbO. We also prepared red clay using 10 wt% CdO nanoparticles and compared its attenuation ability with the red clay prepared with 10 wt% PbO, Bi2O3, and CdO microparticles. We found that the MAC of the red clay with 10 wt% nano-CdO was higher than the MAC of the clay with microparticle samples. Accordingly, nanoparticles could be a useful way to enhance the shielding ability of current radiation shielding materials.

Decommissioning procedure and induced activation levels, calculations and measurements in an 18 MeV medical cyclotron

The present article describes the decommissioning of a self-shielded 18 MeV medical cyclotron IBA Cyclone 18/9 after 14 years of operation. A Monte Carlo simulation of the possible nuclear reactions was performed in order to plan the decommissioning activities. During the cyclotron dismantling, the activities of the cyclotron components, concrete wall and floor samples were measured. Residual activities were analysed by means of an HPGe detector and liquid scintillation counting, and compared with simulation data. Dosimetry of the staff involved in the decommissioning procedure was monitored by individual TL dosimeters and/or digital dosimeter. The cyclotron component analysis confirmed the presence of gamma and pure beta emitters, 22Na, 54Mn, 60Co, 65Zn, 207Bi, 55Fe, 63Ni at different values of specific activity, depending on the positioning of the sample point and on the alloy of the sampled part. In these components the presence of gamma and pure beta emitters was measured 5 years after the shutdown at levels far above clearance limits as defined by the ‘Recommended radiological protection criteria for the recycling of metals from the dismantling of nuclear installations’ (RP89) guidelines. The simulation, carried out by FLUKA Code (version 2020.0.5) on the cyclotron components, provided good agreement with measurements, with a maximum discrepancy of the same order as the uncertainties. Four engineers of the cyclotron maintenance staff were involved in the dismounting of the hottest components and rigging of the cyclotron in the deposit 6 months after shutdown and two engineers were involved during the drilling phase 3.5 years after shutdown. The measured dose from external exposure of the involved staff was lower than 100 μSv person−1 during the first phase and lower than 20 μSv person−1 during the final drilling phase. Measured doses from intake were negligible. In conclusion, the decommissioning of the 18 MeV cyclotron does not represent a risk for the involved staff, but, due to the presence of long-lived radioisotopes, the cyclotron components are to be treated as low level radioactive waste, and stored in an authorised storage area for at least 25 years after shutdown.

Germanium detector dead time—correction for different energies of gamma-ray photons

Measurement of the high-level activities of isotopes leads to an increase in spectrometer dead time, and this influences quality analysis. This study confirms a correlation between dead time and photon energy, especially in the low-energy region. Paralyzed and nonparalyzed methods of calculating dead time were applied and compared for the short-lived 223Ra isotope in cases of high activity analysis. A method of multi-channel counting was utilized for dead time analysis for gamma spectrometry system with HPGe detector. 223Ra and its decay products 219Rn and 111Pb, as short-lived isotopes, play a significant role in the identification of variabilities in dead times for various energy windows. The relative dead time obtained in this study exceeded a value of 31% and had an influence on degradation in quality and quantity analyses. Therefore, this effect should be controlled and corrected.

Gamma radiation shielding properties of recycled polyvinyl chloride composites reinforced with micro/nano-structured PbO and CuO particles

In this work, the γ-ray shielding characteristics of fabricated recycled polyvinyl chloride (rPVC) loaded with micro and nano fillers of PbO and CuO composites were investigated. The PbO/rPVC and CuO/rPVC composites were prepared by changing the percentage of injected materials (PbO and CuO) between 10% and 40% with additions by 10%. Gamma-ray attenuation properties, such as mass attenuation coefficients, linear attenuation coefficients and half-value layer for PbO and CuO-supported polymers were obtained in the energy range between 59.5 and 1408.0 keV using narrow beam transmission geometry. These experiments were carried out by using a HPGe detector and five standard radioactive sources [241Am, 133Ba, 137Cs, 60Co and 152Eu]. Moreover, equivalent atomic numbers and the exposure buildup factor (EBF) values for incident photon energy in the range from 0.015 MeV to 15 MeV were also calculated by Geometric Progression (G-P) method. The experimental results revealed that, the measured values of the linear attenuation coefficients of the PbO/rPVC and CuO/rPVC composites showed their dependence on the type and the concentration of the filler (either PbO or CuO) and also on the incident gamma ray energy. Increasing the filler weight fractions wt.% in the rPVC matrix, increases the linear attenuation coefficients of the composite and the improvement is more significant in case of PbO compared to CuO of the same filler wt.% at the same energy. The results also demonstrated an enhancement in the radiation protection behavior of rPVC composites due to the addition of PbO and CuO fillers as nano-sized particles.

Source and LVis based coincidence summing correction in HPGe gamma-ray spectrometry

The activity of gamma-ray emitting nuclides is calculated assuming that each gamma-ray is detected individually; thus, the magnitude of the coincidence summing signal must be considered during activity calculations. Here, the correction factor for the coincidence summing effect was calculated, and the detection efficiencies of two HPGe detectors were compared. The CANBERRA Inc. GC4018 high-purity Ge detector provided an estimate for the peak-to-total ratio using a point source to determine the coincidence summing correction factor. The ORTEC Inc. GEM60 high-purity Ge detector uses EFFTRAN in LVis to obtain the parameters of the detector and source model and the gamma-gamma and gamma-X match estimates, in order to determine the coincidence summing correction factor. Nuclide analyses, radioactivity comparisons, and analyses of reference material samples were performed utilizing certified reference materials to accurately determine the detection efficiencies. For both Co-60 and Y-88, the detection efficiency for a point source increased by an average of at least 12–13%, whereas the detection efficiency determined using LVis increased by an average of at least 13–15%. The calculated radioactivity values of the certified reference material and reference material samples were accurate to within 3% and 6% of the measured values, respectively.

Gamma radiation shielding performance of CuxAg(1-x)-alloys: Experimental, theoretical and simulation results

Different types of photon shielding parameters such as total mass attenuation coefficient (μ/ρ), linear attenuation coefficients (μ), half value layers (HVL), mean free paths (MFP), effective atomic numbers (ZEff), energy absorption build-up factors (EABF), exposure build-up factors (EBF) and kerma relative to air were investigated for the fabricated Cu–Ag based alloys. The considered parameters were measured through gamma spectrometer equipped with HPGe detector in order to obtain the experimental attenuation coefficients and other related parameters at various photon energy in the energy range 59.5–1332.5 keV. The measured μ/ρ values were confirmed with WinXCOM database results. FLUKA and GEANT4 simulation codes were used to examine the compatibility of the experimental and WinXCOM database results with these simulation codes. The exposure buildup factors of the alloy samples were estimated with help of Geometric Progression fitting formula over photon energy 0.015–15 MeV up to 40 mfp penetration depth. The results revealed that the exhibited effectiveness of Cu0.2Ag0.8 alloys against high energetic photon radiations had a good performance than that of alternative absorbers such conventional concretes, glasses and some alloys. The results of the present survey can be quite useful for possible applications of such materials, especially in nuclear laboratory and reactor core design for preference of effective photon shielding materials.

Measurement of (n,α) and (n,2n) reaction cross sections at a neutron energy of 14.92 ± 0.02 MeV for potassium and copper with uncertainty propagation * *For the flawless operation of the PURNIMA neutron generator during the experiment, we gladly acknowledged the assistance of scientific officers (SO's) of Purnima laboratory and Analytical Chemistry Division, BARC, Mumbai, India. For the financial assistance

Experimentally measured neutron activation cross sections are presented for the (n,α) , (n,α) , and (n,2n) reactions with detailed uncertainty propagation. The neutron cross sections were measured at an incident energy of 14.92 ± 0.02 MeV, and the neutrons were based on the t(d,n)α fusion reaction. The (n,α) reaction was used as a reference reaction for the normalization of the neutron flux. The pre-calibrated lead-shielded HPGe detector was used to detect the residues' γ-ray spectra. The data from the measured cross sections are compared to the previously measured cross sections from the EXFOR database, theoretically calculated cross sections using the TALYS and EMPIRE codes, and evaluated nuclear data.

The cross section measurements of 131X produced in 238U(n,f) reaction induced by 14 MeV neutron

The cross sections of 131Sb and 131gTe produced from the 238U(n,f) reaction around the neutron energy of 14 MeV were measured by using γ-ray spectrometry with a low background HPGe detector. During the irradiating time, the neutron flux of T(d,n)4He reaction was monitored by following the α-particle. The cross-section ratio of 90Zr(n,2n)89Zr to 93Nb(n,2n)92mNb reaction was used in the determination of neutron energy varied with the emission angle. In the process of calculating results, the Al was used as monitor to eliminate the influence of neutron flux. The obtained fission cross-sections at the neutron energies of 14.1 ± 0.3, 14.5 ± 0.3, and 14.7 ± 0.3 MeV are 31.11 ± 1.72, 29.63 ± 1.80, and 25.56 ± 1.71mb for 238U(n,f)131Sb reaction and 21.82 ± 1.05, 20.40 ± 1.07, and 17.05 ± 0.98 mb for 238U(n,f)131gTe reaction, respectively.

Investigation of the effect of copper contact pin on efficiency in HPGe detectors using Monte Carlo method

The full energy peak efficiency (FEPE) determined by experimental or Monte Carlo (MC) simulation methods is a very important parameter in HPGe detectors. Since FEPE depends on the detector's geometric parameters, the parameters provided by the manufacturer are of great importance in modeling the detector with the MC method. The most important reason for the discrepancy between MC and experimental calculations is the lack of accurate information about the detector's geometric properties. The thickness of the copper contact pin in the middle of the detector hole is not given by the manufacturer. In this study, the effect of copper contact pin thickness on detector efficiency was investigated by using the PHITS 3.24 MC simulation program both at different copper contact pin radii and at different detector-source distances. The efficiency values were calculated for photons in the energy range of 59.5 keV-1408 keV, at 4 different distances, namely 5 cm, 13.25 cm, 15 cm, and 20 cm and for the radii of copper contact pins increased from 1 mm to 3.5 mm at 0.5 mm intervals. According to the results, it has been determined that the presence of copper contact pins causes a change in detector efficiency up to 1.9%, especially in the high energy region, and has no effect on the detector efficiency in the low energy region. In addition, it has been observed that the effect of copper contact pin thickness on detector efficiency is almost independent of the source-detector distance.

Activation cross sections of some neutron-induced reactions in the energy range of 13.82–14.71 MeV

Abstract The cross sections of the reactions 70Zn(n,2n)69mZn,74Ge(n,α)71mZn and 90Zr(n,2n)89m,gZr in the energy range of 13.82–14.71 MeV were measured by the activation technique in combination with HPGe detector γ-ray spectroscopy. The measured results were compared with other experimental data and with the data given in the library TENDL-2019. A nuclear model calculation based on the code TALYS-1.8 was also performed after adjustment of an input parameter for the spin distribution of level density. The results of this work strengthen the database and could be useful in further evaluation of the data.

The Radiation Shielding Properties of the Low Temperature Alloys

Solder alloys are significant for many applications. In addition to the physical electrical and magnetic properties of solders, it is of great importance to determine the shielding properties of the materials in order to prevent the harmful effects of radiation on human health if they are used in systems containing radiation. In this study, the shielding properties of some low temperature solder alloys were investigated. The alloy samples were tested by the gamma-ray emitted from 241Am radioactive source 59.5 keV energy settled into HpGe detector. The highest mass attenuation coefficient and the lowest HVL and MFP values were determined for studied samples. The results showed that SDO-4 with the lowest shielding parameters, it is much more suitable for the radiation shielding among the studied sample.

Determination of radioactivity levels in feldspathic dental ceramics

Activity concentrations of 42 different feldspathic dental ceramic powders were determined using a gamma spectrometer with an HPGe detector. The average 238U, 232Th and 226Ra activity concentrations of the specimens were 126 ± 8 Bq kg−1, 5.6 ± 0.5 Bq kg−1 and 12.7 ± 1.2 Bq kg−1, respectively. The average 40K activity was found as 2855 ± 89 Bq kg−1 ranging from 2252 ± 70 Bq kg−1 to 3522 ± 110 Bq kg−1 due to high potassium content in dental ceramics. None of the activity concentration measurements exceeded the limits by EC and ISO.

Measurement of the 145Sm half-life

The half-life of 145Sm has been measured by means of the reference source method with a HPGe detector. The long-lived radionuclide 44Ti was mixed into the source for reference. The time-dependency of the 145Sm/44Ti activity ratio was followed by assessing the count-rate ratio of their characteristic gamma-ray emissions at 61.2 keV (145Sm) and 67.9/78.3 keV (44Ti) in spectra recorded over periods of typically one day. In total, 220 measurements were performed over a period of 384 days or about one half-life period. The experiment and ensuing uncertainty budget are discussed in detail. Different error propagation is applied for random uncertainties, autocorrelated structures in the fit residuals, and potential systematic errors. The result for the 145Sm half-life, 345 (16) d, is compatible with the scarce literature values, however the experimental details of the old measurements were barely documented.

Sensitivity analysis towards trace-uranium detection with γ-γ coincidence NAA

We present an improved approach for detection of trace amounts of 238U by means of neutron activation analysis (NAA). The analysis enhancement is obtained by utilizing γ-γ coincidence counting. An empirical method for evaluating the sensitivity gain in the presence of a large source-related background is presented. A comparison of detection limits is made between two counting schemes; namely, counting single gammas using one HPGe detector versus counting coincident gammas using two HPGe detectors placed face-to-face. A data-validated radiation transport model, created in GEANT4 for a single HPGe detector, is extended to handle the two-detector setup. In this counting scheme, a 238U-detection sensitivity enhancement of about a factor of 8 is predicted for a sample of Saint-Gobain G3 Sapphire when compared to the simpler singles counting approach.

Proteomic Biomarker Analysis of Serum from Japanese Field Mice (Apodemus Speciosus) Collected within the Fukushima Difficult-to-return Zone.

The environmental impact of the Fukushima Daiichi nuclear power plant accident is a source of ongoing concern as there is uncertainty regarding the effects of chronic radiation exposure on local plant and animal life from Fukushima-derived radionuclides. In the current study, changes in proteomic biomarker expression due to chronic environmentally-derived radiation exposures was examined in wild field mice. Serum from 10 wild field mice (Apodemus speciosus) native to the Fukushima difficult-to-return zone and from eight wild field mice native to the Soma area (control) were collected. External dose estimations were completed using measurements of ambient radiation levels and calculating 137Cs concentrations in soil. Internal dose was estimated by counting whole mice using an HPGe detector. Age of the mice was estimated using molar wear. Serum was screened using the aptamer-based SOMAscan proteomic assay technology for changes in expression of 1,310 protein analytes. A subset panel of protein biomarkers that demonstrated significant changes in expression between control and exposed mice was determined and analyzed using Ingenuity Pathway Analysis (IPA). Control animals had a calculated lifetime dose range from 0.001 to 0.007 Gy, and exposed animals had a calculated lifetime dose range from 0.01 to 0.64 Gy. No discernable effect of dose rate was seen as relative dose rate correlated with dose for all samples. Detectable values were obtained for all 1,310 proteins included in the SOMAscan assay. Subset panels of proteins demonstrating significant (p < 0.05) changes in expression with either an upregulated or downregulated 1.5-fold change over control were identified for both the sample cohort inclusive of all exposed samples and the sample cohort restricted to samples from animals receiving "low" dose exposures. These panels of proteins from exposed animals were analyzed using IPA, which highlighted changes in key biological pathways related to injury, respiratory, renal, urological, and gastrointestinal disease, and cancer. Significant changes in expression of proteomic biomarkers were seen in the serum of wild field mice that received environmental exposure to Fukushima-derived radionuclides. Our findings demonstrate novel biomarkers of radiation exposure in wildlife within the Fukushima difficult-to-return zone.

Measurement of the 124Xe (n,2n) reaction cross section induced by 14.8 MeV neutron

The measurement of the 124Xe(n, 2n)123Xe reaction cross section at 14.8 MeV neutron has been performed by the activation method. Quasi monoenergetic neutron beams were produced at the Cockcroft Walton Accelerator of China Institute of Atomic Energy, by the T(d, n)4He reaction. A cell made of polymethyl methacrylate (PMMA) was used as the container of 124Xe (99.95% abundance) gas, and the mass of 124Xe in cell was determined by weighing method. Two high-purity thin 93Nb foils were attached to the cell and performed as monitors of the neutron flux determination by 93Nb(n, 2n)92mNb reaction. After the irradiation, the activities of 123Xe and 92mNb were measured by off-line gamma-ray spectrometric technique using a pre-calibrated HPGe detector. As a result, the cross section has been deduced at 14.8 MeV, and the result was compared with the literature data and evaluations.

Lifetime measurements of excited states in Cr55

Excited states in $^{55}\mathrm{Cr}$ have been populated via the fusion-evaporation reaction $^{48}\mathrm{Ca}(^{11}\mathrm{B},p3n){\phantom{\rule{0.16em}{0ex}}}^{55}\mathrm{Cr}$ at a beam energy of 32 MeV. The Cologne plunger device surrounded by a $\ensuremath{\gamma}$-ray detector array was employed to determine lifetimes with the recoil-distance Doppler-shift method. $\ensuremath{\gamma}$ rays were observed by one Ge EUROBALL cluster detector and five HPGe detectors. $\ensuremath{\gamma}\ensuremath{\gamma}$-coincidence data were analyzed with help from the differential decay-curve method, and precise lifetimes for the first excited states were extracted from the $5/{2}^{\ensuremath{-}}\ensuremath{\rightarrow}3/{2}^{\ensuremath{-}}$ and the $9/{2}^{\ensuremath{-}}\ensuremath{\rightarrow}5/{2}^{\ensuremath{-}}$ transitions. Reduced transition strengths $B(\ensuremath{\sigma}\ensuremath{\lambda})$ were determined and compared to shell-model calculations employing four interactions KB3G, FPD6, GXPF1A, and GXPF1Br. The calculations were also performed for the $N=31$ isotonic chain from $^{53}\mathrm{Ti}$ to $^{61}\mathrm{Zn}$. Inspection of the wave functions as well as particle-plus-rotor model calculations allow for a detailed understanding of the excited states and the reduced transition strength in $^{55}\mathrm{Cr}$. The interactions GXPF1A and GXPF1Br reproduce well the experimental findings in $^{55}\mathrm{Cr}$ and in other isotones.