Radioisotope Source Research Articles

Tritium Power Supply Sources Based on AlGaAs/GaAs Heterostructures

The results of the development of a radioisotope power supply source based on a semiconductor (AlxGa1 –xAs/GaAs) β-radiation energy converter and tritium as a radiation source are presented. The efficiency of this converter with a tritium saturated titanium disk, a green luminescent tritium lamp, or tritium gas as a radioisotope radiation source is compared. For a converter based on the Al0.35Ga0.65As/GaAs heterostructure in a capsule with tritium, efficiency η = 5.9% at a maximum specific output electrical power of 0.56 μW/cm2. Due to their long service life, such stand-alone and compact power supply sources can be used in space and subsea technologies, implantable cardiac pacemakers, biosensors, and portable equipment of high reliability.

Investigation of the Temperature Effect on the Output Parameters of Radioisotope Sources of Electricity Based on Double Energy Conversion of Radiative Decay

When developing long-life radioisotope sources of electricity for various applications (space industry, medicine, nano- and microsystem technology, cryptography, and telecommunications), an important problem lies in determining the temperature range of their reliable operation. The effect of negative and positive temperatures in the range from –60 to +60°C on the output parameters of radioisotope electricity sources based on double energy conversion is studied. It is shown that the open-circuit voltage of a radioisotope source changes four times in the indicated temperature ranges. In this case, the highest power delivered to the load is implemented at a temperature of about 0°C. The effect of temperature on all stages of energy conversion is analyzed. The investigations show that a significant decrease in the open-circuit voltage and the shape of the power-output curve are determined by two mechanisms: a decrease in the luminous intensity of radioluminescent sources of light (the temperature quenching of luminescence) and a decrease in the efficiency of photoconverters. Moreover, in the range of negative temperatures (from –60 to 0°C), the decrease in the efficiency of photoconverters is expressed only slightly, and the main contribution to the change in the output parameters of the sources is introduced by the temperature quenching of luminescence in the source of light. In the range of positive temperatures, both processes have a significant effect on reducing the output voltage and power. The developed radioisotope electrical-power sources based on double conversion can be used in electronic equipment operating at lower temperatures, for example, under Far-North conditions.

Design, manufacturing and calibration of the new models of silver GM activation counters

The innovative single and array models of silver Geiger Muller (GM) activation counters were designed to measure the fast neutron fluence. These counters were made of inner and outer neutron moderators (PE) sections with special designs of silver foil and thin wall GM counter. The Monte Carlo N-Particles MCNPX2.6 code was used to simulate the optimum dimension and energy responses of these counters. The two counters were calibrated by means of 241Am-Be radioisotopic neutron source with 16Ci activity. The detection efficiencies of single and array counters were 0.07 ± 0.005 and 0.785 ± 0.05, respectively. The energy responses of these two counters were determined to be flat for low energy neutrons of 2 keV to 1 MeV which were rarely observed in other neutron detectors. The directional sensitivities were also improved after rejecting scattered neutrons by surrounding materials which will be of high importance for intense pulsed neutron sources.

Effect of gamma irradiation on photoluminescence of MEH-PPV/detonation nanodiamond polymer composite

We studied the photoluminescence of MEH-PPV polymer films and a MEH-PPV/DND nanocomposite after exposure to gamma irradiation from a radioisotope source of cesium-137 with doses of 0.5–12.2 kGy for H2O and their relaxation processes after exposure to the maximum dose. Exposure to gamma irradiation leads to the formation of new structural units in MEH-PPV/DND, which are similar in structure to the structural units of the conductive PPV polymer. A microscopic model is proposed that describes the evolution of MEH-PPV/DND photoluminescence in the context of the formation of bonds between aliphatic radicals of polymers and the surface of detonation diamond.

Detector sensitivity and calibration for yield measurements in fusion neutron sources

Abstract In this study, numerical simulations have been carried out to evaluate the effect of moderator parameters (material/dimensions) for standard radio-isotopic/fission (Pu-Be and 252Cf) and fusion sources (D-D). These sources are used for calibrating thermal neutron detectors, which are often employed for determining absolute neutron yield in fusion-based sources. Simulations have been carried out using the Monte Carlo based “FLUKA” code for typical moderator materials and 3He/BF3 proportional counter based thermal neutron detectors. Results have been validated against experimentally measured values for an available radio-isotopic source. A notable difference in the detector efficiency was observed due to the source energy term. The detection efficiency ratio for Pu-Be:D-D:252Cf sources with the same moderator thickness (optimized for 252Cf) was found to be 1 : 1.16±0.02 : 1.40±0.03. For these three types of sources, optimum acrylic (Perspex) and polypropylene moderator thicknesses have been observed to be 100,100,80 mm and 80,80,60 mm respectively with the corresponding efficiency ratio of 1 : 1.15±0.02 : 1.34±0.008.

Optimization of 241Am-Be emission direction in neutron porosity tools for improving the precision in determining the porosity in calcite formation

Investigation of Hydrocarbon reservoir is important, so it is essential to predict and explore them precisely. One of the methods is well logging, which can transfer the probe or tool in the well to measure one or more characteristics. Nuclear well logging includes radioisotope source and at least one detector. In this work, emission direction of neutrons from the 241Am-Be neutron source toward the calcite formation has been investigated using MCNPX 2.6 to obtain the best precision in determining the liquid porosity. The results show that emission direction of 20 in degree with the resolution porosity of 3% in counts is recorded in detectors; as well; there is no decrease in depth penetration, providing speed of tools.

Neutron transport correction and density calculation in the neutron-gamma density logging

Formation density is one of the most important parameters in formation evaluation. The radioisotope chemical sources are widely used in conventional nuclear logging tools. Considering security and environmental risks, the pulsed neutron generators have successfully replaced the Am–Be source in neutron porosity logging tools. However, the Cs-137 is still mainly used for bulk density measurement. Recently, there is a growing interest in measuring bulk density with the pulsed neutron generators. Although the neutron-gamma density (NGD) is also calculated by gamma-ray attenuation, the neutron transport correction is needed for neutron-induced inelastic gamma rays. A new method of NGD calculation is developed based on the coupled-field theory. This new method does not require additional neutron transport correction. Additionally, a structural model of NGD logging tool is designed by Monte Carlo method, which includes two gamma-ray detectors and two epithermal neutron detectors. According to the proposed method, the corrected inelastic gamma-ray count rate response can eliminate the effect of neutron transport and meet the law of gamma-ray attenuation. The Monte Carlo simulation results indicate that the accuracy of NGD measurement is 0.02 g/cm3, which is close to the conventional gamma-gamma density logging (0.015 g/cm3). The NGD measurement is insensitive to porosity and formation fluids in the pore space.

Planar and textured surface optimization for a tritium‐based betavoltaic nuclear battery

Remote, terrestrial, and space sensors require sources that have high enough power and energy densities for continuous operation for multiple decades. Conventional chemical sources have lower energy densities and lifetimes of 10 to 15 years depending on environmental conditions. Betavoltaic (βV) nuclear batteries using β--emitting radioisotopes possess energy densities approximately 1000 times greater than conventional chemical sources. Their electrical power density (Pe,vol in W/cm3) in a given volume is a function of β--flux surface power density , surface interface type between radioisotope and transducer, β- range, and transducer thickness and conversion efficiency (ηs). Tritium is the most viable β--emitting radioisotope because of its commercial availability, low biotoxicity, half-life, and low energy, which minimizes the penetration depth and damage of transducer. To maximize Pe,vol, tritium in solid or liquid form must be used in the βV nuclear battery. A Monte Carlo source model using MCNP6 was developed to maximize the Pe,vol of a tritium-based βV nuclear battery. First, a planar coupling configuration with different tritiated compounds (ie, titanium tritide and tritiated nitroxide) and a semiconductor transducer (4H-SiC) with thicknesses of 1 and 100 μm were modeled. The results showed that β--source efficiency (ηβ), which is the percentage of energy deposited in the transducer, decreased as the tritiated compound's mass density increased. The highest Pe,vol was dependent on a combination of characteristics: specific activity (Am in Ci/g), mass density, and 4H-SiC layer thickness. The tritiated nitroxide with the highest Am at 2372 Ci/g produced the highest Pe,vol at 2.46 mW/cm3. Second, a 3-D coupling configuration was modelled to increase surface interfacing between the radioisotope source and textured transducer surface. 3-D coupling configuration increased the percentage of energy deposited into the transducer because of more surface interfacing between the transducer and source in the same volume. The tritiated nitroxide was selected as the radioisotope source coupled with five different textured surface feature types. The Pe,vol as a function of textured surface feature and gap, where the radioisotope is located, width was calculated for 1- and 100-μm 4H-SiC layer thicknesses. Results showed that ηβ increased compared with planar coupling configuration (ie, approximately 56.2% increase over planar with cylindrical hole array) except with the rectangular pillar array. Still, the rectangular pillar array produced the highest Pe,vol at 4.54 mW/cm3 with an increasing factor of 2.29 compared with the planar coupling configuration.

Simulation and optimization of current generation in gallium phosphide nanowire betavoltaic devices

The geometry of a gallium phosphide nanowire (NW) array has been optimized for maximum current generation in a betavoltaic (BV) device. The energy capture efficiency for various device geometries with different radioisotope source compounds was calculated in GEANT4. A validation of GEANT4 for BV device simulation was performed by comparing a model output with the available bulk semiconductor BV performance data, followed by predictions of the performance of NW-based devices. The pitch and the diameter of the NWs were found to have the most significant impact on the β-generated current density, with the optimum diameter-to-pitch ratios ranging from 0.55 to 0.8, depending on the source. The energy capture efficiency improved when low energy beta (β) emitters were used. For devices utilizing 63Ni source compounds, the β-generated current densities approached 0.95 μA cm−2, representing an improvement by a factor as high as 5.8 compared to planar devices. In the case of 3H source compounds, the generated current density was 3.05 μA cm−2, a factor of 15.5 larger than comparable planar devices. However, NW devices utilizing sources with a higher decay energy, such as 147Pm, did not demonstrate any improvements over planar geometries. Using the results for optimum NW geometries, NW-based or other nanostructured devices could be made to surpass the present commercial BV batteries.

Thermodynamics kinetics of boron carbide under gamma irradiation dose

In this paper, high purity boron carbide samples were irradiated by [Formula: see text]Co gamma radioisotope source (0.27 Gy/s dose rate) with 50, 100, 150 and 200 irradiation hours at room-temperature. The unirradiated and irradiated boron carbide samples were heated from 30[Formula: see text]C to 1000[Formula: see text]C at a heating rate of 5[Formula: see text]C/min under the argon gas atmosphere of flow rate 20 ml/min. Thermogravimetric (TG) and Differential Scanning Calorimetry (DSC) were carried out in order to understand the thermodynamic kinetics of boron carbide samples. The weight kinetics, activation energy and specific heat capacity of the unirradiated and irradiated boron carbide samples were examined in two parts, T [Formula: see text] 650[Formula: see text]C and T [Formula: see text] 650[Formula: see text]C, according to the temperature. The dynamic of quantitative changes in both ranges is different depending on the irradiation time. While the phase transition of unirradiated boron carbide samples occurs at 902[Formula: see text]C, this value shifts upto 940[Formula: see text]C in irradiated samples depending on the irradiation time. The activation energy of the unirradiated boron carbide samples decreased from 214 to 46 J/mol in the result of 200[Formula: see text]h gamma irradiation. The reduction of the activation energy after the irradiation compared to the initial state shows that the dielectric properties of the irradiated boron carbide samples have been improved. After the gamma irradiation, two energy barrier states depending on the absorption dose of samples were formed in the irradiated samples. The first and second energy barriers occurred in 0.56–0.80 and 0.23–0.36 eV energy intervals, respectively. The existence of two energy levels in the irradiated boron carbide indicates that the point defects are at deep levels, close to the valence band.

Use of the Optically Stimulated Luminescence Dating and X-Ray Fluorescence Spectrometry Methods as Tools in Paleoseismology

We have investigated the application of luminescence dating to sediment and pottery samples from a paleoseismological trench excavated in the Gyrtoni Fault, Tyrnavos Basin, Central Greece. The samples were dated following the optically stimulated luminescence (OSL) dating method, using the Riso TL/OSL DA-20 reader. The OSL ages were obtained from chemically purified quartz and a single-aliquot regenerative-dose (SAR) protocol was followed for the equivalent dose (De) determination. Additionally, samples were collected and analyzed with the method of X-ray Fluorescence (XRF) spectrometry, in order to assess their elemental composition. Radioisotope sources (109Cd and 241Am) were used for sample excitation, while X-ray spectra were acquired using a Si(Li) detector coupled with standard electronics. The XRF data were submitted to principal component analysis (PCA). This statistical handling aimed to distinguish from which part of the upthrown fault block scarp-derived colluvium and alluvial deposits, parts of the downthrown block were derived and thus estimate the displacement. The results indicated that both the OSL dating method and the XRF analysis combined with PCA can serve as useful tools for paleoseismological investigations.

THE INVESTIGATION ORE DEPOSIT OF RADIOISOTOPE ELEMENT WITHIN THE MINING SUBSTANCE IN THE WEST TIMOR ISLAND NUSA TENGGARA TIMUR

These research inspect the distribution and the accumulation shape of ore depo-sit anomaly of radioisotope element within the mining substance at Sub-district of Amara-si and East – Middle Kupang West Timor Island. The aims of the research: 1) to localize the distribution wide of radioisotope element within the mining substance, 2) to stake out the target reconstruction or shape of accumulation anomaly ore deposit of radioisotope e-lement within the mining substance, 3) to decide counts range of nuclear radiation and thermal counts in the center region of radioisotope source in the West Timor Island. The methods of research consist of: observation, survey, modeling, analysis, and interpretation. Procedures of research including: 1) to observe for determination the po-tential region and grid plot, 2) to calibrate equipment, 3) to measure background count in around of survey location, 4) to measure nuclear radiation and thermal count in the sur-vey location, 5) to correct field count with background count, 6) two and three dimensional characteristic curve of radiation and thermal count plot, contour, contour slice, and distri-bution profile curve of ore deposit anomaly of radioisotope in the mining substance al-so plot, 7) to determine the depth of ore deposit of radioisotope element within the mi-ning substance and radiation power, 8) to calculate radiation intensity for modeling of tar-get reconstruction, 9) to stake out modeling and target reconstruction of ore deposit of ra-dioisotope in the mining substance, 10) interpretation and conclusion. The results of the research: radiation power to determine radiation intensity is 1183911,452 w. Based on characteristic curve, contour, profile curve, and modeling of o-re deposit anomaly (target reconstruction) of radioisotope element within the mining subs-tance, can be proposed that the wide of ore deposit distribution of radioisotope element in rock mineral 1.5 kilometers x 2.0 kilometers, and the wide of ore deposit center region 200 meters x 300 meters. The shape of ore deposit anomaly of radioisotope element wi-thin the mining substance resemble of oval ball vertically with diameters ± 54 meters hori-zontal, and ± 100 meters vertical on the average depth 22 meters, and the shape on abo-ve surface is dented. The range thermal and radiation counts of radioisotope element in the mining substance revolve between 20.3o C to 50.7o C and 9 counts per minute to 117 counts per minute.

Development of a method for multielemental determination in water by EDXRF with radioisotopic source of 238Pu.

A method for determination of Cr, Fe, Co, Ni, Cu, Zn, Hg and Pb in waters by Energy Dispersive X Ray Fluorescence (EDXRF) was implemented, using a radioisotopic source of 238Pu. For previous concentration was employed a procedure including a coprecipitation step with ammonium pyrrolidinedithiocarbamate (APDC) as quelant agent, the separation of the phases by filtration, the measurement of filter by EDXRF and quantification by a thin layer absolute method. Sensitivity curves for K and L lines were obtained respectively. The sensitivity for most elements was greater by an order of magnitude in the case of measurement with a source of 238Pu instead of 109Cd, which means a considerable decrease in measurement times. The influence of the concentration in the precipitation efficiency was evaluated for each element. In all cases the recoveries are close to 100%, for this reason it can be affirmed that the method of determination of the studied elements is quantitative. Metrological parameters of the method such as trueness, precision, detection limit and uncertainty were calculated. A procedure to calculate the uncertainty of the method was elaborated; the most significant source of uncertainty for the thin layer EDXRF method is associated with the determination of instrumental sensitivities. The error associated with the determination, expressed as expanded uncertainty (in %), varied from 15.4% for low element concentrations (2.5-5 μg/L) to 5.4% for the higher concentration range (20-25 μg/L).

Effect of annealing temperature on K-shell X-ray fluorescence parameters of zinc oxide thin films prepared by the sol-gel method

Zinc oxide thin films were grown on a glass substrate by a sol-gel process using a spin-coating technique. The obtained thin films were annealed between 350 °C and 550 °C in 50 °C steps and were then characterized using X-ray diffraction, scanning electron microscopy, and X-ray fluorescence techniques. The samples were stimulated by 59.5 keV gamma rays emitted from an Americium-241 annular radioisotope source. K X-rays emitted by samples were counted using an ultra-low energy germanium detector with a resolution of 150 eV at 5.96 keV. It was found that there was generally a decrease in both the Kβ/Kα X-ray intensity ratios and the K X-ray fluorescence cross sections for zinc oxide between 350 °C and 500 °C, but not at 550 °C. In addition, the X-ray diffraction patterns of the films showed that the transition phase from an amorphous to a polycrystalline hexagonal wurtzite structure was complete at an annealing temperature of 500 °C. The results show that variations in these parameters can be explained by the reorganization of atoms and the charge transfer process due to the effect of the annealing temperature on the elements forming the compounds.

Small Autonomous kW-Level Power Generation Based on Radioisotope and Renewable Energy Sources for the Arctic Zone and the Far East

In the near future, the demand for autonomous kW-level electric power generating plants in the Arctic zone and in the Far East will number in the tens and hundreds. In the present article, the operational characteristics of radioisotope thermoelectric generators with direct conversion of heat into electricity, which are being decommissioned, are analyzed. Schemes for autonomous electric power plants based on renewable sources of energy and electrochemical cells are proposed as alternatives. The problems arising in the development of such plants are analyzed. The results achieved in meeting concrete orders from operating organizations are discussed. It is concluded that autonomous electric power plants with renewable sources of energy and fuel cells hold promise.

Formula omitted]-particle spectrometer based on xenon gas ionization chamber using coplanar electrodes

A parallel plate gas ionization chamber with coplanar electrodes was designed and developed to be used as an α-particle spectrometer measuring low level α-emission rates in environmental samples from the vicinity of nuclear power plants. The chamber was filled with a 3 atm gas mixture of Xe + 0.7% CH4. Its performance was investigated using a mixed α-particle source (237Np, 241Am, and 244Cm). An excellent energy resolution of 59 keV (1.0%) in fwhm for 5.81 MeV 244Cm α-particles was obtained, which is enough to identify major sources of alpha radioisotopes (238U, 235U, 237Np, 239Pu, 238Pu, 241Am, 244Cm) to be found near nuclear power plants.

Тритиевые источники электропитания на основе гетероструктур AlGaAs/GaAs

We report on the development of a radioisotope power source based on the AlхGaAs1-х/GaAs semiconductor converter of β-radiation and tritium as a radiation source. The transducer efficiencies are compared for three types of β-radiation sources: (i) the tritiated titanium disk, (ii) the tritium lamp of green luminescent glow, and (iii) gaseous tritium. When using a converter based on the Al0.35Ga0.65As/GaAs heterostructure in a tritium capsule, the efficiency η = 5.9% is found with the maximum output specific electric power of 0.56 µW/cm2. Due to the prolonged operational life, such autonomous and compact power sources can be utilized in space technology, underwater systems, growth implants, biomedical sensors, and portable mobile equipment of high realibility.

Determination of the flux of a neutron radioisotope source through gold activation method

Introduction: In recent years, the use of neutrons has increased in various fields such as medicine and industry. One of the important applications of neutrons is in medical research. Neutrons have greater linear energy transfer than gamma rays and are considered more dangerous radiation. Hence, the detection and determination of neutron dose is a very important and vital issue. For determination of neutron dose, measurement of its intensity or flux is essential. One of the ways of neutron flux determination is using of gold foil activation method. In this method, the activity of the irradiated foil is proportional to the neutron flux. The gold foil is activated by reaction 197Au (n,γ) 198Au because of the thermal neutron capture,. In the decay of activated gold, a gamma radiation of 411.8 keV is released with a probability of more than 95%, that By measuring the area below the peak associated with this energy, by the NaI(Tl) detector, the activity of gold and then the neutron flux is determined. Materials and Methods: In this paper, an 18 carat gold foil was used to measure the flux of a neutron radioisotope source Ra-Be in the nuclear laboratory of the Faculty of Science of Guilan University. Results: After irradiation, the gold foil was counted by NaI(Tl) detector and its flux was determined Conclusion: Finally, by comparing the measured flux and the flux in the Ra-Be neutron source catalogue, it is resulted that the mentioned method has a good accuracy in determination of the flux of the neutron source.

Investigation of neutron flux by thermoluminescence dosimeters in the neutron-gamma mixed field

Introduction: Neutrons have many applications in various fields, such as medicine. The important application of neutron in medical science is in Boron Neutron Capture Therapy (BNCT). Because of this, determination of neutron dose and flux is critical for the health maintaining of workers and patients exposed to this beam. The neutron radioisotope sources produce gamma rays in addition to the neutrons, and the neutron fields around these sources are mixed-field neutron-gamma fields. Since the separation of the dose of each radiation component in the compound field is difficult, the gamma-neutron mixed field dosimetry is a major challenge. International Commission on Radiation and Units (ICRU) recommends the use of a dosimeter pair for mixed-field dosimetry that have different responses to each component of radiation. One of the effective methods of neutron dosimetry is the use of the TLD-600/700 dosimeter pair. TLD-600 is 6Li enriched and TLD-700 is 7Li enriched. These dosimeters have the same sensitivity to gamma radiations but TLD-600 is sensitive to thermal neutrons because of high capture cross-section of 6Li, while in low fluxes, TLD-700 has no sensitivity to thermal neutrons. Materials and Methods: In this paper, in order to measure the neutron flux of the Ra-Be source in the core laboratory of the Faculty of Science of the University of Guilan. Six TLD-600/700 dosimeters were used in three intervals of 7, 14 and 20 cm from the source Results: After 24 hours of irradiation, the intensity of the thermoluminescence of TLDs was measured by TLD reader. Finally, by calculating the calibration coefficient, the measured flux values by the TLDs were determined and compared with the calculated fluxes Conclusion: Results show that the measured and calculated values are in good agreement.

Demonstration of a Tritiated Nitroxide Nuclear Battery

Unattended, compact, terrestrial and space sensors require sources that have high energy and power densities to continuously operate for 3 to 99 years depending on application. Currently, chemical sources cannot fully satisfy these applications, especially in solid state form. Betavoltaic (βV) nuclear batteries using β--emitting radioisotopes possess energy densities 1000 times greater than conventional chemical sources. Their power density is a function of β- flux saturation point relative to the planar (2D) configuration, β- emission range, and the semiconductor converter, the betavoltaic (βV) cell, properties. The figure of merit is the beta (β-)-flux surface power density (Pβ− in μWn per cm2 footprint), where an optimal portion of incident beta particles penetrates the surrounding semiconductor depletion region. Tritiated nitroxides are favorable radioisotope sources with the potential to have the highest specific activity (Am in Ci/g) and Pβ− for an organic compound in solid form. The goal of this research is to demonstrate a tritiated nitroxide nuclear battery using the planar (2D) coupling configuration. The reproducible tritiation procedure produced stable product with a Am of approximately 635 Ci/g, which was 70% of the theoretical Am. For the nuclear battery demonstration, the tritiated nitroxide, dissolved in methanol, was deposited on a 4H-SiC βV and InGaP photovoltaic (PV) cell using a dispensing apparatus and micropipette. Both devices’ characteristics were measured beforehand using a controlled electron beam source to approximate the surface radioactivity from the deposited radioisotope. The maximum power point (MPP) of the 4H-SiC and InGaP were 7.77 nW/cm2 and 1.63 nW/cm2 with 100 mCi and 67 mCi, respectively. The power and total efficiency were lower than expected due to partial solvent evaporation and droplet thickness. Numerical models using MCNP6 Monte Carlo code were used to simulate an optimal nuclear battery prototype. The models’ accuracy was confirmed with the device calibration curves and a previous metal tritide model based on empirical results. Based on optimal model results, the tritiated nitroxide saturation layer thickness (D0.99) and Pβ−(D0.99) were 10 µm and 558 nW/cm2, respectively, using a 4H-SiC.