scholarly journals RESEARCH PROCESSES OF GAMMA RADIATION DETECTOR FOR DEVELOPING A PORTABLE DIGITAL SPECTROMETER

2020 ◽  
pp. 5-12 ◽  
Author(s):  
O.V. Banzak ◽  
O.V. Sieliykov ◽  
M.V. Olenev ◽  
S.V. Dobrovolskaya ◽  
O.I. Konovalenko
Keyword(s):  
Gamma Radiation ◽  
Dose Rate ◽  
Power Plants ◽  
Gamma Ray ◽  
Dynamic Range ◽  
Radiation Detector ◽  
Minimum Energy ◽  
Sampling Rate ◽  
Nuclear Materials ◽  
Wide Range

When considering methods of combating the illicit circulation of nuclear materials, it is necessary to detect trace amounts of materials, and in many cases not to seize them immediately, but to establish the place of storage, processing, routes of movement, etc. As a result, there is a new demand for isotope identification measurements to meet a wide range of different requirements. Measurements should be carried out in the field in a short time, when results need to be obtained within tens of seconds. The devices with which the personnel work should be small and low-background. Such requirements appear when working to identify cases of illegal trade in nuclear materials and radioactive sources, as well as when solving radiation protection problems and when handling radioactive devices and waste. In this work, new generation radiation sensors and measuring systems based on them have been created, which open up previously unknown possibilities in solving problems of nuclear fuel analysis, increasing the accuracy and efficiency of monitoring technological parameters and the state of protective barriers in nuclear power plants, and creating means for IAEA inspections. For the first time a portable digital gamma-ray spectrometer for radiation reconnaissance in the field was developed and created. Distinctive features of such devices are: The analysis showed that the required value of error due to energy dependence of the sensitivity can be achieved using, for example, Analog Devices 10-bit AD9411 ADCs with a sampling rate of 170 MHz. The number of quantization levels is determined by the requirement to measure the dose rate of gamma radiation with an energy of at least 10 keV. This minimum energy corresponds to the use of 10-bit ADCs. On the basis of the developed model, an ionizing radiation detector for dosimetry was created. Its fundamental difference from known devices is the use of CdZnTe crystals as a primary gamma-ray converter (sensor). The advantages of such a solution, proved by previous studies, made it possible to create a detector with: high resolution, no more than 40 keV; a wider dynamic range of values of the recorded radiation dose rate - from background to emergency operating modes of the reactor; lower value of the energy equivalent of noise.

scholarly journals Compact lightweight imager of both gamma rays and neutrons based on a pixelated stilbene scintillator coupled to a silicon photomultiplier array

2021 ◽  
Vol 11 (1) ◽  
Author(s):  
Jihwan Boo ◽  
Mark D. Hammig ◽  
Manhee Jeong
Keyword(s):  
Gamma Rays ◽  
Gamma Ray ◽  
Sampling Rate ◽  
High Sensitivity ◽  
Pulse Shape Discrimination ◽  
Nuclear Materials ◽  
Silicon Photomultiplier ◽  
Analog To Digital ◽  
Particle Imaging ◽  
Imaging Plane

AbstractDual particle imaging, in which both neutrons and gamma-rays in the environment can be individually characterized, is particularly attractive for monitoring mixed radiation emitters such as special nuclear materials (SNM). Effective SNM localization and detection benefits from high instrument sensitivity so that real-time imaging or imaging with a limited number of acquired events is enabled. For portable applications, one also desires a dual particle imager (DPI) that is readily deployable. We have developed a hand-held type DPI equipped with a pixelated stilbene-silicon photomultiplier (SiPM) array module and low sampling-rate analog-to-digital converters (ADCs) processed via a multiplexed readout. The stilbene-SiPM array (12 × 12 pixels) is capable of effectively performing pulse shape discrimination (PSD) between gamma-ray and neutron events and neutron/gamma-ray source localization on the imaging plane, as demonstrated with 252Cf neutron/gamma and 137Cs gamma-ray sources. The low sampling rate ADCs connected to the stilbene-SiPM array module result in a compact instrument with high sensitivity that provides a gamma-ray image of a 137Cs source, producing 6.4 μR/h at 1 m, in less than 69 s. A neutron image for a 3.5 × 105 n/s 252Cf source can also be obtained in less than 6 min at 1 m from the center of the system. The instrument images successfully with field of view of 50° and provides angular resolution of 6.8°.

Beta Backscattering and Gamma Radiation Absorption Characteristics of Carbon Nanoparticles Contained Concrete Composite

2017 ◽  
Vol 17 ◽  
pp. 31-36 ◽  
Author(s):  
B. Gopal Krishna ◽  
Pooja Prasad ◽  
Vibha Sahu ◽  
Jyoti Prabha Sahu ◽  
Akansha Agarwal
Keyword(s):  
Gamma Radiation ◽  
Carbon Nanoparticles ◽  
Power Plants ◽  
Gamma Ray ◽  
Nuclear Research ◽  
Radiation Absorption ◽  
Particle Accelerators ◽  
Building Structures ◽  
Back Scattering ◽  
Absorption Characteristics

Concrete is a very important composite for making different building structures to absorb different levels of radiation. Nuclear power plants, nuclear research reactors, particle accelerators and linear accelerator in medical institution use concrete in building construction. Nanoparticles or nanocrystals have different properties than their bulk counterparts. The gamma radiation absorption characteristics and beta back scattering by nanoparticles is also different than their counterparts. In this paper, carbon nanoparticles are mixed in the concrete composite during its preparation. The concrete composite with carbon nanoparticles as admixture were analyzed to provide radiation protection. The gamma radiation absorption characteristics and beta back scattering in ordinary and carbon nanoparticles contained concretes have been studied by GM counter. The results show that using carbon nanoparticles as an admixture in the concrete is one of the solutions for gamma ray shielding and beta back scattering. Therefore, it is good to use carbon nanoparticles as admixtures in concrete composites for beta and gamma radiation scattering and absorption respectively.

Experimental Studies of Radiation-Protective Properties of a Modified Titanium Hydride

2022 ◽  
Vol 1049 ◽  
pp. 174-179
Author(s):  
A.A. Karnauhov ◽  
R.N. Yastrebinskii
Keyword(s):  
Gamma Radiation ◽  
Dose Rate ◽  
Titanium Hydride ◽  
Nuclear Power ◽  
Power Plants ◽  
High Efficiency ◽  
Experimental Studies ◽  
Point Sources ◽  
Relaxation Length ◽  
Protective Properties

The results of experimental studies of the protective properties of titanium hydride with respect to neutron and gamma radiation in order to determine the optimal conditions for their use in the composition of the structural radiation protection of the nuclear reactor are presented. The weakening of the basic functionals in the thickness of protection, including the density of fast, intermediate and thermal neutrons, and the dose rate of gamma radiation is established. The functions of weakening the density of neutron flow and the dose rate of gamma radiation are measured in the conditions of "barrier" geometry. Determination of the protective properties of the structure was carried out when the modified titanium hydride fraction was placed in aluminum containers with a filling coefficient of a volume of container 0.63. The relaxation lengths for all neutron groups are close and on average are 9.8 cm. The functions of weakening the dose rate of gamma radiation of point sources Cs-137 and Co-60 are exponential. The weakening of radiation occurs with a constant relaxation length. For energy 0.661 MeV, the relaxation length is 7.1 cm, for energy 1.25 MeV, the relaxation length is equal to 10.1 cm. On the basis of the experimental studies, the high efficiency of the modified fraction of titanium hydride was confirmed during its use in protecting nuclear power plants.

A Simple and Efficient Charge Injection Error Compensation Structure for MOS Sampling Switches

2018 ◽  
Vol 27 (08) ◽  
pp. 1850130 ◽  
Author(s):  
Saeed Naghavi ◽  
Mojde Nematzade ◽  
Niloofar Sharifi ◽  
Tohid Moradi Khanshan ◽  
Adib Abrishamifar ◽  
...  
Keyword(s):  
Input Signal ◽  
Dynamic Range ◽  
Supply Voltage ◽  
Sampling Rate ◽  
Charge Injection ◽  
Sampled Data ◽  
Sampling Errors ◽  
Wide Range ◽  
Sinusoidal Input ◽  
Simulation Results

This paper introduces a new technique to design an analog MOS switch to be used in sampled-data circuits. In any sampled-data system, the accuracy of the sampling switch is a critical parameter to determine the overall performance of the system. To satisfy accuracy requirements of the switch, a novel technique to reduce channel charge injection error is proposed. The proposed switch has a very simple structure and it uses a small area of the chip. Also, it has a low on-resistance and its variation over the input signal range is acceptable. In order to evaluate the performance of the proposed switch, simulations are done in a 0.18[Formula: see text][Formula: see text]m standard CMOS technology. Simulation results show that the sampling errors produced by the channel charge injection is eliminated through a cancellation technique using an auxiliary transistor. The output error charge due to charge injection over a wide range of the input signal variation is very low (less than 1.45[Formula: see text]fC). Also, simulation results show that the proposed switch achieves signal-to-noise plus distortion ratio (SNDR) of 85.05[Formula: see text]dB, effective number of bits (ENOB) of 13.83, total harmonic distortion (THD) of [Formula: see text]87.23[Formula: see text]dB and spurious-free dynamic range (SFDR) of 88.14[Formula: see text]dB for a 1[Formula: see text]MHz sinusoidal input of 800[Formula: see text]mV peak-to-peak amplitude at 50[Formula: see text]MHz sampling rate with a 1.8[Formula: see text]V supply voltage.

scholarly journals Hand-held Dual-particle Imager Implemented with a Multiplexed Low Sampling-rate Readout of a SiPM-based Pixelated Stilbene Array

2020 ◽  
Author(s):  
Jihwan Boo ◽  
Mark Hammig ◽  
Manhee Jeong
Keyword(s):  
Gamma Ray ◽  
Sampling Rate ◽  
High Sensitivity ◽  
Pulse Shape Discrimination ◽  
Nuclear Materials ◽  
Silicon Photomultiplier ◽  
Analog To Digital ◽  
252Cf Source ◽  
Particle Imaging ◽  
Imaging Plane

Abstract Dual particle imaging, in which both neutrons and gamma-rays in the environment can be individually characterized, is particularly attractive for monitoring mixed radiation emitters such as special nuclear materials (SNM). Typical dual particle imagers (DPIs) are not readily deployable and easily portable for hand-held applications because they are implemented using bulky single-crystal scintillators and photomultiplier tubes (PMTs) implemented with a 1:1 channel readout. Effective SNM localization and detection also benefits from high instrument sensitivity so that real-time imaging or imaging with a limited number of acquired events is enabled. We have developed a hand-held type DPI equipped with a pixelated stilbene-silicon photomultiplier (SiPM) array module and low sampling-rate analog-to-digital converters (ADCs) processed via a multiplexed readout. The stilbene-SiPM array (12 × 12 pixels) is capable of effectively performing pulse shape discrimination (PSD) between gamma-ray and neutron events and neutron/gamma-ray source localization on the imaging plane, as demonstrated with 252Cf neutron/gamma and 137Cs gamma-ray sources. The low sampling rate ADCs connected to the stilbene-SiPM array module result in a compact instrument with high sensitivity that provides a gamma-ray image of a 137Cs source, producing 6.4 μR/h at 1 m, in less than 69 seconds. A neutron image for a 3.5 × 105 n/s 252Cf source can also be obtained in less than 6 minutes at 1 m from the center of the system. The instrument images successfully with field of view of 50° and provides angular resolution of 6.8°.

scholarly journals IMPREGNATED SEMICONDUCTOR SCINTILLATOR

2008 ◽  
Vol 18 (04) ◽  
pp. 973-982 ◽  
Author(s):  
SERGE LURYI
Keyword(s):  
Gamma Radiation ◽  
Gamma Ray ◽  
Diffusion Length ◽  
Radiation Detector ◽  
The Body ◽  
Linear Dimensions ◽  
Wide Gap

A semiconductor scintillation-type gamma radiation detector is discussed in which the gamma-ray absorbing semiconductor body is impregnated with multiple small direct-gap semiconductor inclusions of bandgap slightly narrower than that of the body. If the typical distance between them is smaller than the diffusion length of carriers in the body material, the photo-generated electrons and holes will recombine inside the impregnations and produce scintillating radiation to which the wide-gap body is essentially transparent. In this way it is possible to implement a semiconductor scintillator of linear dimensions exceeding 10 cm.

INVESTIGATION OF XENON-FILLED TUBE BREAKDOWN VOLTAGE AND DELAY RESPONSE AS POSSIBLE DOSIMETRIC PARAMETERS FOR SMALL GAMMA RAY AIR KERMA RATES

2020 ◽  
Vol 190 (1) ◽  
pp. 84-89
Author(s):  
Milić Pejović ◽  
Emilija Živanović ◽  
Miloš Živanović
Keyword(s):  
Gamma Radiation ◽  
Dose Rate ◽  
Breakdown Voltage ◽  
Radiation Dosimetry ◽  
Gamma Ray ◽  
Physical Processes ◽  
Rate Range ◽  
Air Kerma ◽  
Delay Response ◽  
Air Kerma Rate

Abstract This paper presents experimental results of dynamic breakdown voltage and delay response as functions of gamma ray air kerma rate for xenon-filled tube at 2.7 mbar pressure. Gamma ray air kerma rate range was considered from 123 nGy h–1 up to 12.3 mGy h–1 in order to investigate the possibility of the application of this tube in gamma radiation dosimetry. It was shown that the variations of the above-mentioned parameters are considerable up to the dose rate of 1.23 μGy h–1, which points to the possibility for application in small dose rate gamma ray dosimetry. Physical processes that make dominant impact to dynamic breakdown voltage and delay response during xenon-filled tube irradiation are also discussed in the paper.

scholarly journals Overview of ion chromatographic applications for the analysis of nuclear materials: Case studies

2021 ◽  
Vol 40 (1) ◽  
pp. 204-219
Author(s):  
Harshala Parab ◽  
Jayshree Ramkumar ◽  
Ayushi Dudwadkar ◽  
Sangita D. Kumar
Keyword(s):  
Dynamic Range ◽  
Simultaneous Detection ◽  
Nuclear Industry ◽  
Nuclear Materials ◽  
Analytical Monitoring ◽  
Wide Range ◽  
Anions And Cations ◽  
Materials Used ◽  
Control Rods

Abstract Accurate, precise, and rapid analytical monitoring of various nuclear materials is essential for the smooth functioning of nuclear reactors. Ion chromatography (IC) has emerged as an effective analytical tool for simultaneous detection of different ions in a wide range of materials used in the nuclear industry. The major advantages over other techniques include superior selectivity and sensitivity for detection of anions and cations, wide dynamic range, and speciation studies of ions. This article provides an overview of different ion chromatographic methodologies developed for the analyses of various nuclear materials such as fuel, control rods, moderator, coolant, and process streams. Comparison of various analytical aspects of IC over the other routine techniques reveals the ease and multidimensional capability of the technique. An insight is given to the modern variations in the field such as coupling of IC with other techniques for the characterization of nuclear matrices, implementation of capillary IC in terms of miniaturization, and so on. The information presented herein will serve as a very useful resource for investigators in the field of characterization of nuclear materials.

scholarly journals APPLICATION OF THE SPECTROMETRIC METHOD FOR CALCULATING THE DOSE RATE FOR CREATING CALIBRATION HIGHLY SENSITIVE INSTRUMENTS BASED ON SCINTILLATION DETECTION UNITS

2017 ◽  
Vol 8 (3) ◽  
pp. 246-253 ◽  
Author(s):  
R. V. Lukashevich ◽  
G. A. Fokov
Keyword(s):  
Monte Carlo ◽  
Gamma Radiation ◽  
Dose Rate ◽  
Integral Transformation ◽  
Photon Radiation ◽  
Detector Response ◽  
Spectrometric Method ◽  
Scintillation Detection ◽  
Highly Sensitive ◽  
Wide Range

Devices based on scintillation detector are highly sensitive to photon radiation and are widely used to measure the environment dose rate. Modernization of the measuring path to minimize the error in measuring the response of the detector to gamma radiation has already reached its technological ceiling and does not give the proper effect. More promising for this purpose are new methods of processing the obtained spectrometric information. The purpose of this work is the development of highly sensitive instruments based on scintillation detection units using a spectrometric method for calculating dose rate.In this paper we consider the spectrometric method of dosimetry of gamma radiation based on the transformation of the measured instrumental spectrum. Using predetermined or measured functions of the detector response to the action of gamma radiation of a given energy and flux density, a certain function of the energy G(E) is determined. Using this function as the core of the integral transformation from the field to dose characteristic, it is possible to obtain the dose value directly from the current instrumentation spectrum. Applying the function G(E) to the energy distribution of the fluence of photon radiation in the environment, the total dose rate can be determined without information on the distribution of radioisotopes in the environment.To determine G(E) by Monte-Carlo method instrumental response function of the scintillator detector to monoenergetic photon radiation sources as well as other characteristics are calculated. Then the whole full-scale energy range is divided into energy ranges for which the function G(E) is calculated using a linear interpolation.Spectrometric method for dose calculation using the function G(E), which allows the use of scintillation detection units for a wide range of dosimetry applications is considered in the article. As well as describes the method of calculating this function by using Monte-Carlo methods and the features of its application. The results of the calculation function G(E) for the detection unit on the basis of NaI(Tl) detector (Ø40 mm, h = 40 mm) to use it as a comparator for kerma rate in the air certification of low intenseе photon radiation fields. 

Sign in / Sign up

Export Citation Format

Share Document