scholarly journals Natural Gamma-Ray Background Characterization in Pyhasalmi Mine

2021 ◽  
Author(s):  
Vladimir Gostilo ◽  
Serhii Pohuliai ◽  
Alexander Sokolov ◽  
Jari Joutsenvaara ◽  
Julia Puputti
Keyword(s):  
Energy Range ◽  
Count Rate ◽  
Gamma Ray ◽  
Background Level ◽  
Nitrogen Gas ◽  
Background Count ◽  
Background Count Rate ◽  
Natural Gamma ◽  
Hpge Spectrometer ◽  
Background Gamma Radiation

<p>We present the results of measuring the gamma-ray background performance of Pyhasalmi mine, the deepest one in Europe. Two underground facilities in Lab 2 (1440 m) and Lab 5 (1410 m) were investigated. Based on measurements made in Lab 2 with a low-background HPGe spectrometer, we determined the integral gamma-ray background count rate in the energy range of 40 keV to 2.7 MeV to be 0.095 s<sup>–1</sup> kg<sup>–1</sup>. The minimum detectable activities of some natural and artificial nuclides were less than 0.071 Bq/kg (<sup>226</sup>Ra), 0.77 Bq/kg (<sup>40</sup>K) and 0.012 Bq/kg (<sup>137</sup>Cs). The specific activities of natural nuclides in the shotcrete covering the walls of the Lab 2 were higher than those in the rock: 100.3 Bq/kg (<sup>232</sup>Th), 161.7 Bq/kg (<sup>226</sup>Ra) and 1171 Bq/kg (<sup>40</sup>K) in the shotcrete covering and 47.6 Bq/kg (<sup>232</sup>Th), 83.1 Bq/kg (<sup>226</sup>Ra) and 1513 Bq/kg (<sup>40</sup>K) in the rock. The measurements showed that the gamma-ray background level in Lab 5 is significantly lower than that in Lab 2. The integrated gamma-ray background count rate for the energy range of 40 keV to 2.7 MeV was 0.028 s<sup>–1</sup> kg<sup>–1</sup> for Lab 5. Purging the measuring chamber of the gamma spectrometer with nitrogen gas at a rate of 0.15 L/h allowed to further improve this parameter to 0.021 s<sup>–1</sup> kg<sup>–1</sup>. In general, the results of this study confirm that the level and energy spectrum of background gamma radiation in the underground facility within the studied energy range is defined mainly by the composition of the walls of the Labs.</p><p> </p>

scholarly journals Newly Designed 0.8-ML Teflon® Vial for Microvolume Radiocarbon Dating

Radiocarbon ◽  
1995 ◽  
Vol 37 (2) ◽  
pp. 743-747 ◽  
Author(s):  
Michael Buzinny ◽  
Vadim Skripkin
Keyword(s):  
Significant Influence ◽  
Count Rate ◽  
Radiocarbon Dating ◽  
Figure Of Merit ◽  
Liquid Scintillation ◽  
Detection Efficiency ◽  
Operational Parameters ◽  
Counting Period ◽  
Background Count ◽  
Background Count Rate

We have tested two versions of an 0.8-ml volume Teflon® vial, designed specifically for radiocarbon dating in the microvolume range, using a modern Quantulus 1220™ liquid scintillation (LS) spectrometer. We determined the counting performance of each vial type in conjunction with different designs of copper holder, viz., with and without the incorporation of a “Teflon® light coupler”. We also compared the losses of sample benzene during a typical 28-day counting period. Results show that neither vial design nor the type of vial holder used in the intercomparison had a significant influence on counting performance. We recorded an absolute 14C detection efficiency of 82% against a background count rate of 0.1 cpm, i.e., a “figure of merit (FM) value” = 67,240. This compares favorably with the operational parameters anticipated for microvolume 14C dating by modern LS spectrometry. However, variations in the design of the sealing method used between the vial types was reflected in an apparent approximate tenfold difference in the amount of benzene lost during routine counting. In the better case, the evaporation loss was equivalent to 0.032 mg of benzene per day.

scholarly journals Precision Measurements of the Minimum Detectable Activity of Radionuclides by Plastic Scintillators for Radiation Monitoring

2019 ◽  
pp. 3-9
Author(s):  
B. Grynyov ◽  
N. Gurdzhian ◽  
O. Zelenskaya ◽  
V. Lyubynskiy ◽  
N. Molchanova ◽  
...  
Keyword(s):  
Count Rate ◽  
Statistical Processing ◽  
Sensitivity Coefficient ◽  
Minimum Detectable Activity ◽  
National Academy Of Sciences ◽  
Background Count ◽  
Background Count Rate ◽  
Energy Radiation ◽  
Detectable Activity ◽  
Plastic Scintillators

This work is devoted to the precision evaluation of the minimum detectable activity (MDA) measurement of gamma radiation radionuc­lides 137Cs and 60Co using plastic scintillators (PS). Scintillators are part of the detectors, which are sensitive elements of radiation portal moni­tors (PM). PS were produced at the Institute for scintillation materials of the National Academy of Sciences of Ukraine (ISMA NAS of Ukraine). The evaluation was generate in accordance with State standard of Ukraine ISO 5725-2:2005 [1] through statistical processing of experi­mental material. This made it possible to take into account the factors that determine the scattering of results without a model equation. Influence of PS sizes, energy radiation detected and distance from PS to the source of ionizing radiation (SIR) on the value of MDA precision measurement was studied. Precision coefficients of radionuclides MDA measurement evalua­ted when changing PS sizes from 500×300×50 mm (7500 cm3) up to 500×500×50 mm (12500 cm3), changing the energy radiation detected from 662 keV (137Cs) until 1332 keV (60Co), changing the distance from the PS to SIR from 10 to 50 cm. The results are presen­ted in Fig. 6—9 and Table. Found, that the greatest contribution to the increase in MDA precision coefficient measurements makes a change in the irradia­tion geometry. It is shown that when distance from PS to SIR increases from 10 to 50 cm, precision values increase: when registering 137Cs, from 3,9 to 7,0 % (V = 7500 cm3) and from 3,6 to 6,3 % (V = 12500 cm3); when registering 60Co, from 3,1 to 4,3 % (V = 7500 cm3) and from 3,2 to 3,8 % (V = 12500 cm3). It is established, that the obtained results are due to the change in the contribution of background count rate to PS count rate under different measurements conditions. This contribution affects on the value of sensitivity coefficient and consequently on the value of MDA and precision coefficient.

scholarly journals Stability of a New, Multichannel, Low-Level Liquid Scintillation Counter System, Kvartett

Radiocarbon ◽  
1995 ◽  
Vol 37 (2) ◽  
pp. 727-736 ◽  
Author(s):  
Sigurđur Einarsson ◽  
Páll Theodórsson
Keyword(s):  
Count Rate ◽  
Liquid Scintillation ◽  
Scintillation Counter ◽  
Counting System ◽  
Background Count ◽  
Background Count Rate ◽  
Low Level ◽  
Quality Control Standard ◽  
Conventional Systems ◽  
Sample Vial

Kvartett is a new liquid scintillation counting (LSC) system for radiocarbon dating that takes a radical departure from conventional systems to obtain a compact, low-level counting system measuring four samples simultaneously. Each sample vial, inside the well of a large NaI(Tl) guard-counter crystal (facing down), sits on top of a vertical PMT. The fourfold counting capacity can be used to increase the number of samples being dated or to get higher precision. The increased throughput helps to keep a rigid quality-control standard. We monitored the background count rate almost continuously for 7 months, and measured the count rate of a standard repeatedly for 2 months. The results show the background and system reproducibility to be stable.

Comparison of a Solid-State Si Detector to a Conventional Scintillation Detector-Monochromator System in X-Ray Powder Diffraction Analysis

1989 ◽  
Vol 4 (3) ◽  
pp. 137-143 ◽  
Author(s):  
David L. Bish ◽  
Steve J. Chipera
Keyword(s):  
Solid State ◽  
Powder Diffraction ◽  
Scintillation Detector ◽  
Count Rate ◽  
Detection Limits ◽  
Background Count ◽  
Background Count Rate ◽  
X Ray ◽  
Graphite Monochromator ◽  
Lower Background

AbstractA new Peltier-cooled solid-state Si(Li) detector has been compared to a traditional scintillation detector/diffracted-beam graphite monochromator system in conventional X-ray powder diffraction applications. Parameters studied included absolute count rates, detector linearity, peak-to-background ratios, detection limits, fluorescent radiation elimination, and peak profile shapes. Comparisons were performed on a Siemens D-500 θ-2θ diffractometer using constant sample and nondetector instrumental parameters. Advantages of the Si(Li) detector include a significantly increased count rate (3.4 - 3.8 times), primarily due to the elimination of the graphite monochromator, slightly lower background count rates, and the ability to change the analysis energy quickly. The higher count rate and slightly lower background count rate of the Si(Li) detector allow collection of data more rapidly than possible with a scintillation detector/diffracted-beam monochromator system and yield improved peak-to-background ratios and detection limits. Significant disadvantages of the Si(Li) detector include pronounced deviation from linearity at low count rates, making accurate measurement of even moderate countrate peaks difficult, and detector shutdown due to 100% deadtime between 4 and 5 × 10 4 counts/s (cps). The Si(Li) detector and the scintillation detector/diffracted-beam monochromator system are comparable in terms of fluorescence radiation elimination, resolution, and peak shape, although it appears tfiat die diffracted-beam monochromator measurably reduces the low-angle portion of the half width of all reflections.

scholarly journals Improving gross count gamma-ray logging in uranium mining with the NGRS probe

2018 ◽  
Vol 170 ◽  
pp. 05001
Author(s):  
C. Carasco ◽  
B. Pérot ◽  
J.-L. Ma ◽  
H. Toubon ◽  
A. Dubille-Auchère
Keyword(s):  
Count Rate ◽  
Gamma Ray ◽  
Uranium Concentration ◽  
Weight Fraction ◽  
Computer Code ◽  
Calibration Coefficient ◽  
Rock Density ◽  
Natural Gamma ◽  
Concrete Blocks ◽  
Parametric Studies

AREVA Mines and the Nuclear Measurement Laboratory of CEA Cadarache are collaborating to improve the sensitivity and precision of uranium concentration measurement by means of gamma ray logging. The determination of uranium concentration in boreholes is performed with the Natural Gamma Ray Sonde (NGRS) based on a NaI(Tl) scintillation detector. The total gamma count rate is converted into uranium concentration using a calibration coefficient measured in concrete blocks with known uranium concentration in the AREVA Mines calibration facility located in Bessines, France. Until now, to take into account gamma attenuation in a variety of boreholes diameters, tubing materials, diameters and thicknesses, filling fluid densities and compositions, a semi-empirical formula was used to correct the calibration coefficient measured in Bessines facility. In this work, we propose to use Monte Carlo simulations to improve gamma attenuation corrections. To this purpose, the NGRS probe and the calibration measurements in the standard concrete blocks have been modeled with MCNP computer code. The calibration coefficient determined by simulation, 5.3 s-1.ppmU-1 ± 10%, is in good agreement with the one measured in Bessines, 5.2 s-1.ppmU-1. Based on the validated MCNP model, several parametric studies have been performed. For instance, the rock density and chemical composition proved to have a limited impact on the calibration coefficient. However, gamma self-absorption in uranium leads to a nonlinear relationship between count rate and uranium concentration beyond approximately 1% of uranium weight fraction, the underestimation of the uranium content reaching more than a factor 2.5 for a 50 % uranium weight fraction. Next steps will concern parametric studies with different tubing materials, diameters and thicknesses, as well as different borehole filling fluids representative of real measurement conditions.

Development of Waste Monitor of Clearance Level to Ensure Social Reliance on Recycled Metal From Nuclear Facilities

2003 ◽  
Author(s):  
Takatoshi Hattori ◽  
Michiya Sasaki
Keyword(s):  
Gamma Rays ◽  
Count Rate ◽  
Gamma Ray ◽  
Surface Contamination ◽  
Contamination Level ◽  
Background Count ◽  
Beta Particles ◽  
Radioactivity Level ◽  
Metal Waste ◽  
Clearance Level

Metal and concrete wastes in the decommissioning of nuclear facilities are classified according to their radioactivity level after decontamination. Radioactive waste below the clearance level (e.g., 0.4Bq.g−1 for Co-60 in Japan) can be disposed of as general industrial waste or recycled. Metal wastes mainly originate from equipment in buildings, except for the metal bars in reinforced concrete. Since contaminated equipment must be decontaminated after dismantling, the main target of measurement would be fragments of equipment, of various shapes, numbers and sizes. In order to transport such metallic fragments out of controlled areas, a surface contamination survey must be performed to confirm that the contamination level is below the legal standard level (e.g., 4Bq.cm−2 for beta or gamma emitters in Japan) in addition to satisfying the clearance level. Taking account of social reliance on recycled metal after inspection of the clearance level and the surface contamination level, it is important to remove the possibility of overlooking contamination above these levels in the recycled metal. The measurement of beta rays is suitable for determining surface contamination on metal because almost none of the beta particles from inside the metal can be detected and the detected radiation can be mostly limited to that from the surface. This is the reason why a survey meter for measuring surface contamination has a detector with a higher sensitivity for beta particles than for gamma rays. Considering the characteristics of the survey meter, it may be difficult to measure the contamination level of the surface of a metal fragment, particularly when the surface is not flat. Moreover, in the case of internal contamination of a small metal pipe, measurement is impossible. The permeability of gamma rays is much greater than that of beta particles. Therefore, gamma rays can be detected even from internal contamination in metal. For gamma ray measurement, accurate and easy calibration of the actual radioactivity level and count rate obtained using a measurement instrument is important. If gamma ray measurement can confirm that the radioactivity level is less than about 400Bq, both the clearance level and the surface contamination level could be inspected simultaneously. In addition, the great amount of labor needed for manual inspection using a survey meter could be saved, and there will be no possibility of missing hot spots of radioactivity due to human error. In this study, a new technique for precise and automatic measurement of gamma emitters in metal waste has been developed using 3D noncontact shape measurement and Monte-Carlo calculation techniques to objectively confirm that the specific radioactivity level of metal waste satisfies the clearance level and furthermore, that the surface contamination level of the metal waste is below the legal standard level. The technique can yield a calibration factor for every measurement target automatically and realizes automatic correction for the reduction of the background count rate in gamma measurements due to the self-shielding effect of the measurement target. A practical monitor (Clearance Automatic Laser Inspection System, CLALIS) has been developed. The accuracy of the automatic calibration and correction of background reduction of the practical monitor has been clarified using mock metal wastes of various shapes, numbers and sizes. It was found that the values measured using the present monitor and the actual radioactivity level agreed within +/−20%, and the corrected and actual background reductions agreed within +/−2%. The detection limit of the present monitor was estimated as being 100Bq for Co-60, taking into consideration the calibration error and correction error of the reduction of the background count rate. The monitor accomplished precise measurements with a 100sec (30sec for gamma ray measurement, 30sec for background measurement) process time per inspection. This indicates that approximately 5 tons of metal waste can be measured per day (1,000 tons per year) in 20kg batches at that process speed.

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