Gamma-ray Scintillation Spectrometers Research Articles

Determination of neutron moderation parameters through neutron captured gamma ray emission – A novel method

The neutron moderation properties such as relative thermal lethargy, average logarithmic energy decrement and average number of collisions for elemental and composite materials were determined experimentally by using (n, γ) reaction. The Am-Be source used in the experiment emits neutrons of energies from thermal to fast with an average energy of 4.2 MeV along with the emission of 4.43 MeV gamma radiation due to de-excitation of 12C* nucleus. The 2.2 MeV gamma radiation emitted from paraffin due to capture of thermal neutrons by hydrogen in paraffin and 4.43 MeV gamma radiations emitted due to de-excitation of 12C* nucleus were detected with 2 in. × 2 in. NaI(Tl) gamma ray scintillation spectrometer coupled to multi-channel analyzer. Using the intensity ratio of 2.2 MeV and 4.43 MeV gamma radiations, the neutron moderation parameters have been determined for elemental targets such as C, Al, Fe, Cu, Zn, Pb and for composite materials such as Perspex, Cadmium Acetate dihydrate, Magnesium Acetate tetrahydrate, Cobalt Acetate tetrahydrate, Nickel Acetate tetrahydrate and Brass.

Radiation protection parameters of glasses with different yttria additives at different photon energies

Additive yttria in manufacturing glasses yieldsbetter shielding properties which can be used for solar cell covers. Selected ratios of yttria and other oxides have been added to borate-based glasses for the synthesizing via technique of melt quenching. The radiation shielding and structural properties were changed after the addition of yttria in the studied glasses. These glasses were studied using x-ray diffraction and scintillation gamma-ray spectrometer. The glasses have been exposed to photons with 0.662, 1.173 and 1.33 MeV energies that emitted from 137Cs and 60Co radioactive sources and the photon intensities were detected using a 3 × 3 inch NaI(TI) detector. The mass attenuation coefficients (μm), half value layers (HVL), tenth value layer (TVL), the mean free paths (MFP), effective atomic numbers (Zeff) and effective electron densities (ED) values have been measured experimentally. The results are also compared to theoretical values outcomes by using XCOM. The results indicate that the YABB10 glass containing 10 % yttria has the lowest HVL, TVL and MFP values than the other glasses. It indicates that the use of Y2O3 in glass increases the gamma shielding properties which means that these materials have a promising use as a cover layer protection.

Improvement of mechanical properties and radiation shielding performance of AlBiBO3 glasses using yttria: An experimental investigation

Nuclear radiation shielding and rigidity parameters of the newly developed Al(0.1-x)B0.6Bi1.8O3Y2x glasses were studied to elucidate their potential as gamma radiation shielding material. In this study, bismuth borate glass systems with different amount of yttria additive were synthesized using melt-quenching technique. The material properties such as radiation shielding performance, structural and mechanical behaviors have changed with yttria additive into Bi-borate glasses. To obtain these changes, x-ray diffraction, scintillation gamma-ray spectrometer, and pulse-echo measurements have performed. The results showed that yttria additive has increased the density of glasses as well as radiation shielding performance. Also, the rigidity parameters were investigated not only experimentally but also theoretically. The results showed that G5 sample is superior nuclear radiation shielding sample with lowest HVL and MFP values. Moreover, yttria additive has increased the bridging oxygen atoms, enhances the coordination number as well as the cross-link density. The enhancement of all of these parameters indicates that these glasses are an excellent candidate to be used as a radiation shielding material. It can be concluded that yttria additive for aforementioned glass system has an important impact in mechanical and nuclear shielding properties yet further investigations should be considered for different types of applications.

Determination of effective atomic number of composite materials using backscattered gamma photons – A novel method

The effective atomic number (Zeff) of composite materials has been determined by measuring the backscattered gamma photons at 180°. A 661.6 keV gamma photons from 137Cs radioactive source are allowed to scatter at an angle of 180° from the sample. The backscattered photons at 180° from the sample are detected with “2 × 2″ NaI(Tl) scintillation gamma ray spectrometer coupled to 16 K Multi Channel Analyzer (MCA). It is observed experimentally that the intensity of backscattered photons increases initially with increasing the thickness of the target and then it becomes saturated above a certain thickness. By knowing the saturated thickness values for known atomic number of the elemental targets, the Zeff of composite materials has been determined. The experimentally measured Zeff values have been compared with theoretical values predicted by direct method, power law method and AutoZeff code.

Mapping of radiation anomalies using UAV mini-airborne gamma-ray spectrometry

Localization of size-limited gamma-ray anomalies plays a fundamental role in uranium prospecting and environmental studies. Possibilities of a newly developed mini-airborne gamma-ray spectrometric equipment were tested on a uranium anomaly near the village of Třebsko, Czech Republic. The measurement equipment was based on a scintillation gamma-ray spectrometer specially developed for unmanned aerial vehicles (UAV) mounted on powerful hexacopter. The gamma-ray spectrometer has two 103 cm3 BGO scintillation detectors of relatively high sensitivity. The tested anomaly, which is 80 m by 40 m in size, was investigated by ground gamma-ray spectrometric measurement in a detail rectangular measurement grid. Average uranium concentration is 25 mg/kg eU attaining 700 mg/kg eU locally. The mini-airborne measurement across the anomaly was carried out on three 100 m long parallel profiles at eight flight altitudes from 5 to 40 m above the ground. The resulting 1 s 1024 channel gamma-ray spectra, recorded in counts per second (cps), were processed to concentration units of K, U and Th, while total count (TC) was reported in cps. Increased gamma ray intensity of the anomaly was indicated by mini-airborne measurement at all profiles and altitudes, including the highest altitude of 40 m, at which the recorded intensity is close to the natural radiation background. The reported instrument is able to record data with comparable quality as standard airborne survey, due to relative sensitive detector, lower flight altitude and relatively low flight speed of 1 m/s. The presented experiment brings new experience with using unmanned semi-autonomous aerial vehicles and the latest mini-airborne radiometric instrument. The experiment has demonstrated the instrument's ability to localize size-limited uranium anomalies.

Teaching about natural background radiation

Ambient gamma dose rates in air were measured at different locations (indoors and outdoors) to demonstrate the ubiquitous nature of natural background radiation in the environment and to show that levels vary from one location to another, depending on the underlying geology. The effect of a lead shield on a gamma radiation field was also demonstrated to emphasize the important role of shielding in radiation protection. The measurements were carried out with a Geiger–Muller (GM)-based dosimeter and a NaI scintillation gamma-ray spectrometer, which are normally available in physics laboratories. Radioactivity in household materials was demonstrated using a gas mantle as an example.

Study of the Influence of Forest and Peat Fires on the Radiation Situation in the South-Western Regions of the Bryansk Region

Introduction. The paper considers the influence of radiation forest and peat fires on the spread of radioactive contamination, which affects the well-being of thousands of people. The state of the environment in vast territories is deteriorating; negative socio-economic processes are developing. This is a serious problem of two states: the Russian Federation and the Republic of Belarus. The objectives of the presented work are to study the radiation situation in forests and peat bogs located in the border areas of the Bryansk region, as well as to study the possibility of transferring radioactive materials during forest and peat fires. Materials and Methods. The facts that clarified the theoretical basis of the presented research are highlighted in the scientific literature. The authors took into account, in particular, that: – the activity of radionuclides in the soil decreases in direct proportion to the depth; – a peat fire is an uncontrolled burning; – emissions of caesium-137 fractions (137Cs) during a fire can reach 3–4 %. We know the areas of forests in radiation-contaminated territories (RCT) of the Bryansk region from the applied literature and official sources. The most problematic areas from this point of view have been identified.A mobile radiometric laboratory, a scintillation gamma-ray spectrometer MKS-AT6101S were involved in the expedition research. The results of field gamma-ray spectrometry were recorded in three localities. The calculations for a hypothetical fire were carried out using the SAUR AIUS RSCHS 2030 software tool. Results. The consequences of large and prolonged fires in the exclusion zone of the Chernobyl nuclear power plant are analyzed. It is established that the incidents did not lead to dangerous consequences for the population. The total effective dose of inhalations was ~0.003 % of the permissible level of irradiation. It is noted, however, that the forest soil of the Bryansk region has received significant damage from 137Cs pollution. The density of such pollution exceeded 5 Ci/km2 in 40 % of the affected forests. Of these, an indicator of 15-40 Ci/km2 and more was recorded in 16 %, in some quarters — up to 200 Ci/km2. It is established that zones with a high contamination density (40 Ci/km2) will remain in the region until 2026. Five districts are particularly problematic: Gordeevsky, Zlynkovsky, Klintsovsky, Krasnogorsky and Novozybkovsky. When fixing and predicting harm, the authors of the presented work proceeded from the following fact: during fires, the combustion products (forest litter, grass and undergrowth) contain more radioactive substances than the crowns of trees. In this regard, measurements were not made at a significant height. The field study route was chosen based on the available data on the maximum level of radioactive contamination. The ambient dose equivalent rate (ADER) of gamma radiation recorded at 2,757 points did not exceed 1.2 μSv/h (with an average value of 0.2-0.3 μSv/h). Measurements in the marshes did not reveal traces of 137Cs at a depth of more than 40 cm. The maximum activity of 137Cs was observed in the upper (0–4 cm) soil layer (up to 65 %). Taking into account the data obtained, the possibility of radioactive containation in the event of a forest fire was assessed. According to the calculations in the software environment, radioactive contamination will spread to 348 hectares. The density of radioactive contamination of the area may increase by 5–10 % (from the initial one). 33 people will suffer; there is a threat of death of 1 person. It has been established that a burning peat bog is the most powerful and long-term source of radioactive contamination, therefore it is important to prevent peat and forest fires. This will reduce the transfer of radionuclides and emissions of radioactive fumes. Remote and surface radiation monitoring facilities should be developed. Discussion and Conclusion. The registered ADER is not dangerous for the health of the population of the Bryansk region. However, frequent fires significantly increase the likelihood of transferring active 137Cs to residential areas. In this sense, timely monitoring and forecasting of fires is relevant. The authors formulated proposals to improve the technical and technological components of the solution of the considered problem. 1. To clarify the radiation situation, all-terrain vehicles should be equipped with: – means of registering the radiation situation; – software and hardware complex for automatic collection, analysis of information and its fixation in databases. 2. There should be a reliable cellular communication between all rapid response units in the emergency zone. Further research is focused on the creation of fast-deployable radiation monitoring modules and mobile aerial photography complexes using drones in the emergency zone.

Determination of Fe and Zn in biological samples by radiochemical neutron activation analysis

Radiochemical NAA methods have been developed for the simultaneous determination of Fe and Zn in biological samples. The method involves reactor irradiation, dissolution in 3M HCl and solvent extraction followed by counting on a scintillation gamma-ray spectrometer. Iron was separated with aqueous cupferron and extracted into chloroform while Zn was extracted with 2-thenoyl trifluoroacetone (TTA) into methyl isobutyl ketone (MIBK). Reaction conditions such as pH and the effect of solvents and various ions were studied using tracer activities. The methods have been employed for trace level determination of Fe and Zn in NBS, SRMs, Bowen's Kale, IAEA CRMs and other plant leaves.

Simultaneous determination of exponential background and gaussian peak functions in gamma ray scintillation spectrometers by maximum likelihood technique

Simultaneous fitting of peaks and background functions from gamma-ray spectrometry using multichannel pulse height analysis is considered. The specific case of Gaussian peak and exponential background is treated in detail with respect to simultaneous estimation of both functions by using a technique which incorporates maximum likelihood method as well as a graphical method. Theoretical expressions for the standard errors of the estimates are also obtained. The technique is demonstrated for two experimental data sets.

A compton-suppressed low level gamma-ray scintillation spectrometer for the samples of extended geometry

A Compton-suppressed gamma-ray scintillation spectrometer having high sensitivity and selectivity has been described. The spectrometer is intended to be used for the identification and estimation of low levels of gamma radioactivity in a broad range of sample sizes containing a moderately compelx mixture of gamma-emiting radionuclides. The spectrometer uses a 7.6 cm dia.×76. cm thick Nal(TI) detector having anticoincidence guard of a rectangular plastic scintillator block of size 38 cm×38 cm×30 cm placed inside u 15 cm thick lead shield. Various details of the setup are given and the use of such a system for the analysis of complex spectra of low level activity has been discussed. The anticoincidence guard provides a gross background reduction by a factor of 2.8 for the energy interval between 50 keV and 1600 keV. Performance evaluation of the system with respect to background reduction. Compton suppression factor, peak/total ratio and peak prominance factor have been discussed.

Multiple linear regression analysis scintillation gamma-ray spectra: Theoretical and practical considerations

Application of the method of multiple linear regression (MLR) as a data-analysis technique for gamma-ray scintillation spectrometer data requires knowledge of (1) the response matrix of the spectrometer and (2) the covariance matrix of the unknown spectrum (or spectra). Neither matrix is known exactly. The response matrix can usually be determined with good precision and an estimate of the covariance matrix can usually be made. Perturbation or errors in either matrix, however, will introduce biases in the results. In addition, in the formulation of the method, assumptions are made and criteria are specified which may be appropriate for particular applications. Once generated, the particular formulation is frequently assumed to be generally applicable; additional biases can thereby be introduced. In this report, theoretical and practical considerations are delineated which are important to the satisfactory utilization of the MLR method. Subjective judgments which are introduced in the formulation of the method are indicated, and in view of these judgements, specific areas are described where care must be exercised, both in the acquisition of the data and in the subsequent processing.

A Compton-suppressed coincidence gamma-ray scintillation spectrometer with large NaI(Tl) crystals

A Compton-suppressed coincidence gamma-ray scintillation spectrometer has been assembled to obtain high sensitivity and selectivity in gamma-ray spectral measurements. The spectrometer is intended to satisfy requirements for the identification and estimation of low levels of gamma-ray radioactivity in a broad range of sample sizes and masses containing moderately complex mixtures of gamma-emitting radionuclides. The spectrometer detector is made of NaI(Tl) and consists of two identical subassemblies, each of which contains a central crystal integrally packaged with, and surrounded by, an annular Compton- suppression crystal. Each central crystal is nominally 6 sol1 2 ″ in dia. by 4″ thick, and is optically coupled to a 2″ thick crystal of unactivated sodium iodide (NaI). The annular crystal is 13 1 2 ″ in dia. by 6 1 2 ″ in length, with a nominal 6 1 2 ″ dia. hole through the center. For each subassembly, the exact diameters of the central crystal and of the annular hole are such that a 3 3 2 ″ space exists between the outer surface of the central crystal and the adjacent surface of the annulus. This space is filled with magnesium oxide to optically decouple the central crystal from the annulus. A 9″ dia. by 1 3 4 ″ thick NaI(Tl) backscatter-suppression detector can be optionally included as part of this detection system. Other components of the spectrometer are the following: (1) the photomultiplier assembly; (2) the detector mount and shield; (3) the amplification and gating circuitry; and (4) the pulse height analyzer. The spectrometer detector response in the following modes of operation is presented: (1) gamma sum; (2) gamma-gamma coincidence; (3) gamma-gamma coincidence sum; (4) beta-gated gamma sum. The detector qualities evaluated were resolution, background, degree of Compton-suppression, efficiency, and relative sensitivity.

Decay scheme of Pd111 and Pd111m

The gamma spectra accompanying the decay of the 22 min. Pd111 and the 5.5h Pd111m, were studied with a scintillation gamma ray spectrometer. Seven new gamma transitions were found to be associated with the decay of Pd111. The energies of these gamma rays are 220, 300, 350, 500, 770 and 1150 keV. A scheme representing the decay of Pd111 and Pd111m is proposed.

Decay Of Eu159

The decay scheme of Eu 159 was investigated by using beta and gamma ray scintillation spectrometers with a fast-slow coincidence circuit. The half-life of Eu 159 was measured as 19.0±0.5 min. Beta rays with end-point energies 2.57±0.05, 2.35±0.05, 1.90±0.05, 1.75±0.05, 1.50±0.05 and 1.0±0.1 MeV, and gamma rays with energies 1500, 1100, 800, 670, 220, 145, 130, 90 and 75 keV were found. On the bases of the experimental results obtained by β-γ, γ-β and γ-γ coincidence measurements, a tentative decay scheme of Eu 159 was proposed. The assignments of the excited levels were made based on the Nilsson model and the systematics in neighbouring odd-N nuclei. The systematic behaviour of low excited levels of gadolinium isotopes and N =95 isotones are discussed.

Properties of the Cross Sections for theS32(γ, np)P30,Ca40(γ, np)K38g, andZn66(γ, np)Cu64Reactions from 50 to 300 MeV

Yield values for the ${\mathrm{S}}^{32}(\ensuremath{\gamma}, np){\mathrm{P}}^{30}$, ${\mathrm{Ca}}^{40}(\ensuremath{\gamma}, np){\mathrm{K}}^{38g}$, and the ${\mathrm{Zn}}^{66}(\ensuremath{\gamma}, np){\mathrm{Cu}}^{64}$ reactions have been determined in the 50- to 300-MeV energy range. Targets of natural sulfur, calcium, and zinc were irradiated with bremsstrahlung from the University of Illinois 300-MeV betatron, and the product yields were determined from measurements of their radioactivities with a gamma-ray scintillation spectrometer. In order to extract cross-section information from the yield data, the ${\mathrm{S}}^{32}(\ensuremath{\gamma}, np){\mathrm{P}}^{30}$ reaction yield data from threshold to 90 MeV reported by Bonazzola et al. were normalized to our data and used to complete the yield curves near the reaction thresholds. Cross sections averaged over 30-MeV-wide energy bins have been determined for the ${\mathrm{S}}^{32}(\ensuremath{\gamma}, np){\mathrm{P}}^{30}$ reaction, but it was possible to determine only the behavior of integrated cross sections in the other two cases because of larger statistical uncertainties in the yield data. The previous work on the ${\mathrm{S}}^{32}(\ensuremath{\gamma}, np){\mathrm{P}}^{30}$ reaction at energies below 90 MeV showed that the cross section peaks at a value of 4.3 mb at 28 MeV and then decreases at higher energies. The cross sections resulting from our work reach zero near 100 MeV and then rise, reaching a value of 1.0 mb near 220 MeV. The integrated cross section to 300 MeV is 190\ifmmode\pm\else\textpm\fi{}12 MeV mb. The integrated cross sections to 300 MeV for the ${\mathrm{Ca}}^{40}(\ensuremath{\gamma}, np){\mathrm{K}}^{38g}$ and the ${\mathrm{Zn}}^{66}(\ensuremath{\gamma}, np){\mathrm{Cu}}^{64}$ reactions are 88\ifmmode\pm\else\textpm\fi{}14 MeV mb and 400\ifmmode\pm\else\textpm\fi{}60 MeV mb, respectively. Using the results of Stein et al. on the number of coincident neutron-proton pairs observed from quasideuteron absorption processes in various nuclei, we estimate that events associated with quasideuteron absorption processes account for 25% or less of the observed ($\ensuremath{\gamma}, \mathrm{np}$) cross sections above 140 MeV. A mechanism in which a meson is produced and escapes from the nucleus, followed by a nuclear cascade and nucleon evaporation caused by the recoiling nucleon, can account reasonably well for the remainder of the observed cross sections. However, a sizeable contribution from processes in which a meson is reabsorbed cannot be ruled out, because of the absence of detailed information about the fraction of the meson production events that leads to the ($\ensuremath{\gamma}, \mathrm{np}$) product.

Decay of Eu159

The decay scheme of Eu 159 was investigated by using beta and gamma ray scintillation spectrometers with a fast-slow coincidence circuit. The half-life of Eu 159 was measured as 19.0±0.5 min. Beta rays with end-point energies 2.57±0.05, 2.35±0.05, 1.90±0.05, 1.75±0.05, 1.50±0.05 and 1.0±0.1 MeV, and gamma rays with energies 1500, 1100, 800, 670, 220, 145, 130, 90 and 75 keV were found. On the bases of the experimental results obtained by β-γ, γ-β and γ-γ coincidence measurements, a tentative decay scheme of Eu 159 was proposed. The assignments of the excited levels were made based on the Nilsson model and the systematics in neighbouring odd-N nuclei. The systematic behaviour of low excited levels of gadolinium isotopes and N =95 isotones are discussed.

Internal conversion coefficients of Ni 56

The internal conversion coefficients of the 0.164, 0.272 and 0.484 MeV gamma rays of Ni 56 have been determined utilizing a scintillation gamma ray spectrometer and a Siegbahn-S)ätis intermediate-image beta-ray spectrometer. Respective conversion coefficients of (1.2±0.1) × 10 −2, (3.4±0.2) × 10 −3 and (1.5±0.15) × 10 −3 were found. Comparison with theoretical conversion coefficients expected for pure multipole transitions and use of the Weisskopf single-particle estimates for the gamma-ray transition probabilities lend considerable support to an analysis of the low-excited states of Co 56 in terms of a simple shell model for an odd nucleus.

A system for stabilizing the scintillation spectrometer

A system for stabilizing the gain of a gamma-ray scintillation spectrometer is described. The system utilizes as a reference signal the counting rate from a window discriminator set on a negative slope in the spectrum. A stabilization factor of 120 and a 21 h stability of ± 0.2% were obtained. This system differs from earlier ones by not requiring any modifications in the amplifier system to be stabilized.

A three crystal NaI(Tl) gamma-ray scintillation spectrometer

A three crystal NaI(Tl) gamma ray scintillation spectrometer has been constructed and tested. It is used as a Compton, anti-Compton and as a pair spectrometer. The main advantage of this spectrometer is its special geometry which leads to relatively higher efficiencies for the Compton and pair spectrometers as well as to a simple anti-Compton spectrometer.

Determination of some energy levels in light and medium weight nuclei with a precision gamma ray scintillation spectrometer

A stabilized scintillation spectrometer was used to measure the energy of gamma rays emitted from the following reactor-neutron-produced radioactive sources: Al 28, Cl 38, K 42, Sb 124, Ni 65, Cu 66, Rb 88 and V 52. The energy determinations were made with an accuracy of the order of 1 keV, an improvement of an order of magnitude in several cases.