High-purity Germanium Gamma Spectrometer Research Articles

Zenith angle dependence on cosmic-ray background in HPGe gamma spectrometers by using GEANT4 simulation

In this study, we investigate the impact of zenith angle variations on cosmic-ray induced background in High-Purity Germanium (HPGe) gamma spectrometers using a coincidence technique based on plastic scintillator-Germanium detectors. We utilize an HPGe detector (Model GC2018 Mirion Ge Detector) enclosed within a low-activity cylindrical lead shield (Model 747E Mirion Lead Shield). For cosmic ray detection, a coincidence detection system with plastic scintillator detectors was positioned on top of the lead shielding. The zenith angle at the Germanium detector is computed using the dimensions of the square plastic scintillator and its distance from the Germanium detector center. We carried out measurements of cosmic-ray induced background in an HPGe gamma spectrometer with a square plastic configuration (80cm x 80cm), equivalent to a 45° zenith angle. The experimental measurements were compared with GEANT4 simulation data. The results demonstrate a good agreement between the measured energy spectrum and the simulated data across the energy range of 0.05 to 47 MeV. Further investigations into the effects of varying zenith angles provide valuable insights for optimizing HPGe spectrometer setups with minimized background interference.

Analysis of radon progeny contamination

ABSTRACT In the present work, the radon contamination has been investigated with a high-purity germanium detector and dual ‘gamma + beta’ and ‘alpha’ contamination metre. Radioactive radon (222Rn) and products of its decay were collected by well-known swiping technique. Naturally occurring dust deposited on a plastic surface was collected and the radiation emissions coming from the collected sample were measured and analysed. The main radioactive isotopes contributing to the gamma spectrum were determined. The time behaviour of radon progeny contributing to gamma and alpha activities was investigated and compared with a modelled forecast. The minimal detectable activity when the radon progeny could be determined with a high-purity germanium gamma spectrometer has been established. Also, the minimum activity, measured with a dual contamination metre, has been recognised at which the presence of radon progeny contamination can be determined using gamma spectrometry. A new method is proposed for determining the duration of radon contamination and the time when the contamination occurred. The method is based on the analysis of the relative gamma activities of radon progeny (214Pb and214Bi) measured at different times after the radon contamination.

Cumulative fission yield measurements with 14.7 MeV neutrons on 238U* *Supported by the National Natural Science Foundation of China (11975113)

The fission yield data in the 14 MeV energy neutron induced fission of 238U play an important role in decay heat calculations and generation-IV reactor designs. In order to accurately measure fission product yields (FPYs) of 238U induced by 14 MeV neutrons, the cumulative yields of fission products ranging from 92Sr to 147Nd in the 238U(n, f) reaction with a 14.7 MeV neutron were determined using an off-line γ-ray spectrometric technique. The 14.7 MeV quasi-monoenergetic neutron beam was provided by the K-400 D-T neutron generator at China Academy of Engineering Physics (CAEP). Fission products were measured by a low background high purity germanium gamma spectrometer. The neutron flux was obtained from the 93Nb (n, 2n)92mNb reaction, and the mean neutron energy was calculated using the cross-section ratios for the 90Zr(n, 2n)89Zr and 93Nb(n, 2n)92mNb reactions. With a series of corrections, high precision cumulative yields of 20 fission products were obtained. Our FPYs for the 238U(n, f) reaction at 14.7 MeV were compared with the existing experimental nuclear reaction data and evaluated nuclear data, respectively. The results will be helpful in the design of a generation-IV reactor and the construction of evaluated fission yield databases.

Grass/Soil Concentration Ratios (CRs) of Natural 226Ra, 232Th, and 40K in Vetiveria zizanioides (Vetiver Grass) in Thailand

This work investigated the grass/soil concentration ratios (CRs) of natural radionuclides (226Ra, 232Th, and 40K) in vetiver grass collected from three different regions of Thailand using an HPGe (high-purity germanium) gamma spectrometer. The sampled grass blades were divided into three equal parts to investigate the dependence of grass/soil CRs on the positions of grass blades. The results showed that 40K had the highest grass/soil CRs due to being an essential nutrient for most plants, with ranges of 1.12–7.92, in comparison with 226Ra and 232Th that had ranges of 0.32–1.48 and 0.20–1.31, respectively. In addition, the results indicated that the natural radionuclides investigated were uniformly transferred and accumulated throughout grass blades (p < 0.05), which implied the independence of CRs on the section of the leaf. Furthermore, it was found that the soil samples collected from the central region of Thailand had the highest activity concentrations in all radionuclides, mainly due to a primary composition of dark clay, leading to the highest contents of potassium and calcium (Total K and Total Ca, respectively), and the highest cation exchange capacity (CEC) that led to high concentrations of nutrients and radionuclides. Based on the results from this work, the vetiver grass presented potentials to be utilized as radionuclide absorbers in polluted soil and water due to its high CRs in comparison with other reported plants.

Methods and Influencing Factors for the Simple and Rapid Identification of Depleted Uranium Weapon Use under Battlefield Conditions.

The purpose of this paper is to explore how to rapidly and easily identify depleted uranium (DU) samples under battlefield conditions and to study the factors that influence their measurement. The air-absorbed dose rate and surface contamination levels for DU samples of 2-330 g were measured using a patrol instrument and portable energy spectrometer. The results were analyzed in accordance with IAEA standards for judging radioactive substances. The energy spectra of 5-g quantities of DU samples were analyzed using a high-purity germanium gamma spectrometer, and the uranium content of 100 mg DU samples was determined with an inductively coupled plasma mass spectrometer to clarify the type and composition of the uranium. The same batches of DU samples were identified using a portable gamma-ray spectrometer. We added 0-5 g environmental soil powders at different proportions. After sealing, the spectra were collected with a detection distance of 1-5 cm for 10 min. The activities of U and U nuclides in the samples were detected with an NaI(TI) scintillation detector. The U and U mass abundances in samples were calculated from measured specific activities. The sample was determined to contain DU if the U to U ratio was below 0.00723. It is found that for detecting DU materials with a low activity, surface contamination level measurements are more effective than calculating the air-absorbed external irradiation dose rate. Hence, for low-activity samples suspected to be radioactive, a radiometer with a high sensitivity for surface contamination is recommended, and the optimal measurement distance is 1-3 cm. Under all detection conditions, U can be identified using a portable gamma spectrometer, whereas U can only be detected under certain conditions. If these nuclides can be detected simultaneously, a U to U ratio of below 0.00723 indicates the presence of DU. The main factors affecting this identification include the sample mass, sample purity, measurement distance, and measurement time. For the rapid identification of DU with a portable gamma-ray spectrometer, the mass of uranium in the sample must be more than 1 g, the measuring distance needs to be less than 1 cm, and the measuring time must be 1-10 min. It is feasible to use a portable gamma-ray spectrometer to rapidly identify the types and composition of nuclides in DU samples. The detection of U activity is a precondition for the identification of DU.

Optimization of decay time in radioactivity detection of building ceramics and analysis of radioactivity level of ceramics in Nan’ an city

The basic principle of radioactivity and the technical method for measuring the radioactivity of ceramic tiles with low background multi-channel high-purity germanium gamma spectrometer are introduced. The reasonable optimization of the balance time of tile radioactivity detection has certain reference value for improving the level of radioactivity detection technology. Combined with the data analysis of the special quality supervision and random inspection of ceramic tiles in the past five years, the radioactivity indicators of ceramic tile products in the jurisdiction are all at a qualified level, effectively eliminating the hidden dangers of radioactive safety of ceramic tile products and ensuring the health and safety of the citizens.

Ratio of spectral averaged cross sections measured in standard 252Cf(sf) and 235U(nth,f) neutron fields

The results of systematic evaluations of spectrum averaged cross section (SACS) measurements in the fission neutron fields of 252Cf and 235U are presented. The data form a complete database of high-threshold experimental SACS measured in the same installation under the same conditions and using the same high purity germanium gamma spectrometer. This is crucial to reduce the uncertainty of the ratio and the data scattering and therefore, to minimize discrepancies compared to cross section measured under different conditions in different laboratories. This new dataset complements and extends earlier experimental evaluations. The total emission of the 252Cf neutron source during the experiments varied from 9.5E8 to 4.5E8 neutrons per second. The emission was derived in accordance to the data in the Certificate of Calibration involving absolute flux measurements in a manganese sulphate bath. Concerning 235U fission neutron field, the irradiations were carried out in a specifically designed core assembled in the zero power light water LR-0 reactor. This special core has a well described neutron field. After the irradiation, the low volume irradiated samples to be measured by gamma spectrometry were placed directly on the upper cap of a coaxial high purity germanium (HPGe) detector in a vertical configuration (ORTEC GEM35P4). High volume samples were homogenized and strewn into the Marinelli beaker. The HPGe detector is surrounded by the lead shielding box with a thin inner copper cladding and covered with rubber for suppression of background signal and bremsstrahlung. The experimental reaction rates were derived for irradiated samples from the Net Peak Areas (NPA) measured using the semiconductor HPGe detector. The measured reaction rates are used to derive the spectrum-averaged cross sections. Furthermore, measured reaction rates are also compared with MCNP6 calculations using various nuclear data libraries, in particular IRDFF evaluations.

Power calibration methodology at the CROCUS reactor

In the present article, we detail the method used to experimentally determine the power of the CROCUS zero-power reactor, and to subsequently calibrate its ex-core monitor fission chambers. Knowledge of the reactor power is a mandatory quantity for a safe operation. Furthermore, most experimental research programs rely on absolute fission rates in design and interpretation – for instance, tally normalization of reaction rate studies in dosimetry, or normalization of power spectral density in neutron noise measurements. The minimization of associated uncertainties is only achieved by an accurate power determination method. The main experiment consists in the irradiation, and therefore, the activation of several axially distributed Au-197 foils in the central axis of the core, which activities are measured with a High-Purity Germanium (HPGe) gamma spectrometer. The effective cross sections are determined by MCNP and Serpent Monte Carlo simulations. We quantify the reaction rate of each gold foil, and derive the corresponding fission rate in the reactor. The variance weighted average over the distributed foils then provides a calibration factor for the count rates measured in the fission chambers during the irradiation. We detail the calibration process with minimization of respective uncertainties arising from each sub-step, from power control after reactivity insertion, to the calibration of the HPGe gamma spectrometer. Biases arising from different nuclear data choices are also discussed.

Annual committed effective dosage from natural radionuclides by ingestion of local food growing in mineral mining area, Sri Lanka.

Sri Lanka's largest mineral sand deposit occurs in Pulmoddai, and the surrounding area of the mineral sand deposit has been identified as a high natural background radiation area. The food crops grown in the area are used for human consumption. However, the long-term consumption of high radioactive foodstuff may cause health defects. The objective of the present study was to assess the radiological impact on individuals due to ingestion of foodstuffs grown in Pulmoddai area. Seasonal food crops growing in the area, including cereals, vegetables, nuts, fruits and yams were collected and used to prepare common meal plans consumed by the residents. Samples were analyzed with high-purity germanium gamma spectrometer for activity concentrations. Activity concentration of 40K was observed in higher amounts in every food sample. Trace amounts of 137Cs 232Th and 7Be were also identified. The total committed effective dosage to an adult from gamma-emitting radionuclides from cooked meals was 0.1482mSvyear-1, and raw foodstuff was 0.0667mSvyear-1, which are far below than the harmful levels declared by International Atomic Energy Agency. Results concluded that foodstuff and cooked meals consumed by the people who live in Pulmoddai, Sri Lanka, are radiologically safe.

Phase change microcapsules with lead tungstate shell for gamma radiation shielding and thermal energy storage

Novel microencapsulated phase change materials (MEPCMs) composed of the lead tungstate (PbWO4) shell and paraffin core were designed for shielding of gamma radiation as well as thermal energy storage. Such MEPCMs were prepared via self-assembly methods and in-situ precipitation. The PbWO4 shell with excellent photon attenuation can give the resulting MEPCMs an acceptable gamma radiation shielding capability. The chemical composition and structure of microcapsules samples were studied by X-ray diffractometer (XRD) and Fourier-transform infrared spectroscopy (FTIR). The effects of different core/shell mass ratios on the surface morphology and microstructure of the MEPCMs were determined by scanning electron microscopy (SEM), energy-dispersive spectrometer (EDS), and transmission electronic microscopy (TEM). It was confirmed that the microcapsules exhibit a distinct core-shell structure and a perfect spherical shape. Differential scanning calorimetry (DSC) and thermogravimetric analysis (TGA) were used to present thermal stability and thermal-storage capability of MEPCMs. A high-purity germanium gamma spectrometer to measure the attenuation coefficient of the microcapsules for gamma rays showed that the MECPMs has good γ-rays-shielding property. The multifunction microcapsules in this study have great potential applications for building energy conservation as well as wearable personal protection in nuclear energy engineering.

Rapid estimation of environmental radioactivity surrounding Xiangshan uranium deposits, Jiangxi province, Eastern China

Abstract The surveys of terrestrial gamma dose rate, radon concentration indoor and in water and specific activity of radionuclides of soil were carried out in 14 villages and a town in Xiangshan uranium deposit and surrounding area, Jiangxi province, Eastern China, in 2017-2018, using a scintillator dosemeter, an ionization chamber and a high-purity germanium gamma spectrometer to study radiation status in these places after remediation. A radioactive hot spot was discovered in a village near the mining office, where specific activity of 238U, 226Ra, 232Th and 137Cs of soil was as high as 1433 ± 76 Bq/kg, 1210 ± 62 Bq/kg, 236 ± 13 Bq/kg and 17 ± 1.1 Bq/kg, respectively. The dose rate on a waste rock heap was about 2423 nGy/h. Approximately 50% of the houses in a village near the uranium mining site had radon concentrations that exceeded 160 Bq/m3. There was a significant positive correlation between indoor radon concentration and outdoor gamma dose rate (R2 = 0.7876). The abnormal radon concentration was observed in a rising spring sample providing residents with tap water up to 127.1 Bq/l. Four tap water samples and three of five well water samples exceeded the limit of radon concentration of drinking water in China (11.1 Bq/l). The mean annual effective doses from gamma dose rate data were 0.86 mSv/y and 1.13 mSv/y for indoor radon. The study shows that there are some radioactively contaminated places surrounding the Xiangshan uranium mine. The local outdoor dose rate averages may be used to estimate local indoor radon concentrations.

Automated separation of short-lived 72Ga from fresh fission products based on tandem column chromatography

A method for automated separation of short-lived activation product of 72Ga from fresh fission products has been developed. Based on the ingenious media selection, optimum column chromatography conditions investigation, multi-column (di-(2-ethylhexyl)-phosphoric acid, tri-n-butyl phosphate, aluminum oxide, activated carbon) combination, and process automation design, the proposed method provides a better recovery of 89 % for gallium and excellent decontamination factors of more than 104 for fission products even for large sample volume. Due to effective elimination of matrix interferences, this method coupled with a well-type high-purity germanium gamma spectrometer realized highly selective and sensitive determination of 72Ga in spiked complex soil samples.

Feasibility studies on the burnup measurement of fuel pebbles with HPGe gamma spectrometer

The feasibility of utilizing a High Purity Germanium (HPGe) detector for the fuel element burnup measurement in a future Modular Pebble Bed Reactor (MPBR) was studied. First, the HPGe spectrometer was set-up for running the detector at high count rates while keeping the energy resolution adequately high to discriminate the Cs-137 peak from other interfering peaks. Based on these settings, the geometrical conditions are settled. Next, experiments were performed with Co-60 and Cs-137 sources to mimic the counting rates in real applications. With the aid of KORIGEN and MCNP/G4 simulations, it was demonstrated that the uncertainty of the Cs-137 counting rate can be well controlled within 3.5%. Finally, a full size prototype was tested in comparison with detailed Monte Carlo simulation and the efficiency transfer method was further utilized for efficiency calibration. To reduce the uncertainty in the efficiency transfer process, a standard point source embedded in a graphite sphere was used for efficiency calibration. The correction factor due to pebble self-attenuation was carefully studied.