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ORIGINAL ARTICLE
Year : 2017  |  Volume : 40  |  Issue : 2  |  Page : 95-98  

Improvement in minimum detectable activity for low energy gamma by optimization in counting geometry


Health Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai, Maharashtra, India

Date of Submission14-Mar-2017
Date of Decision06-Apr-2017
Date of Acceptance24-Apr-2017
Date of Web Publication13-Jul-2017

Correspondence Address:
Pradyumna Lenka
Health Physics Division, Bhabha Atomic Research Centre, Trombay, Mumbai - 400 085, Maharashtra
India
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Source of Support: None, Conflict of Interest: None


DOI: 10.4103/rpe.RPE_12_17

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  Abstract 

Gamma spectrometry for environmental samples of low specific activities demands low minimum detection levels of measurement. An attempt has been made to lower the gamma detection level of measurement by optimizing the sample geometry, without compromising on the sample size. Gamma energy of 50–200 keV range was chosen for the study, since low energy gamma photons suffer the most self-attenuation within matrix. The simulation study was carried out using MCNP based software “EffCalcMC” for silica matrix and cylindrical geometries. A volume of 250 ml sample geometry of 9 cm diameter is optimized as the best suitable geometry for use, against the in-practice 7 cm diameter geometry of same volume. An increase in efficiency of 10%–23% was observed for the 50–200 keV gamma energy range and a corresponding lower minimum detectable activity of 9%–20% could be achieved for the same.

Keywords: EffCalcMC, geometry optimization, high purity germanium, minimum detectable activity, self-attenuation


How to cite this article:
Gupta A, Lenka P, Sahoo S K, Kale P K, Ravi P M, Tripathi RM. Improvement in minimum detectable activity for low energy gamma by optimization in counting geometry. Radiat Prot Environ 2017;40:95-8

How to cite this URL:
Gupta A, Lenka P, Sahoo S K, Kale P K, Ravi P M, Tripathi RM. Improvement in minimum detectable activity for low energy gamma by optimization in counting geometry. Radiat Prot Environ [serial online] 2017 [cited 2019 Sep 15];40:95-8. Available from: http://www.rpe.org.in/text.asp?2017/40/2/95/210573


  Introduction Top


Gamma spectrometry is one of the most widely used nondestructive technique for estimation of gamma emitting radionuclides in environmental samples. Qualitative estimation (identification of radionuclide) is a simple process and is performed by energy calibration of the system, where the characteristic gamma energies of the analyte are used for identification of the radionuclide. However, quantitative measurement of the gamma emitters present in the sample can be done by efficiency calibration, which involves numerous parameters to be considered while doing so. Unlike alpha and beta detectors, efficiency of gamma detectors varies with photon energy and is also dependent on the matrix, sample to detector distance and geometry of the sample. Environmental samples constitute low concentrations of many radionuclides and sample constituents vary widely for different matrices. Factors such as active detector volume, gamma abundance are invariant for a given detector system. Factors like time of counting and sample weight can be varied for better precision of measurement, however with limitations. With a view to this fact, it is imperative to optimize the sample geometry to maximize the measurement efficiency and to minimize the self-attenuation of low gamma energies within the matrix for a given sample size.

Radionuclide estimations through gamma spectrometry involve measurement of many low energy gamma photons (e.g. 59.5 keV of 241Am, 63.3 keV of 234Th, 143.8 keV of 235U etc.), which may suffer significant amount of attenuation in the sample matrix itself. This affects the photon efficiency and may result poor precision of measurement.[1] Furthermore, the minimum detection level for the corresponding photon energy gets affected. Most environmental samples contain very low level of natural and manmade radionuclides. Therefore, a low detection level is desirable for gamma spectrometric analysis of environmental samples. The minimum detectable activity (MDA) is calculated from the formula,[2]



Where B: Background counts in T seconds; γ: Gamma abundance (%) and W: Weight of the sample (kg); and ∈: Photon efficiency (%) which gets affected by a change in geometry.

Most laboratories use mainly two kinds of geometries for calibration and sample analysis; cylindrical and marinelli beaker geometries. Marinelli beaker offers better counting efficiency than cylindrical container of the same volume because of its geometry around the detector. However, these containers are difficult to manufacture in small volumes and are used mainly for samples of very small concentrations of analyte, where a larger volume of samples need to be analyzed. Therefore, cylindrical containers are the most common geometry used for environmental sample analysis such as soil, sediment, ashed biological samples, etc.

In this study, a high purity germanium (HPGe) detector system (p–type, R.E. 50%) was used to determine the optimum cylindrical geometry to achieve the maximum detection efficiencies for gamma energies ranging from 50 to 3000 keV using a Montecarlo based commercial software “EffCalcMC.” In this paper, the observations for low energy range 50–200 keV for different cylindrical geometries and the corresponding optimum counting geometry for the lowest attainable MDA is suggested.


  Materials and Methods Top


A p–type co axial HPGe detector of 50% relative efficiency was used for the study. The detector is shielded by 10 cm thick lead rings with graded shielding and is coupled with an 8K multichannel analyzer and spectrum analysis software (SpectraLineGP) for spectrum acquisition and interpretation. The system is calibrated for energy and efficiency using IAEA standard reference materials [3] and has the resolution of 0.8 keV at 122 keV of 57Co and 1.8 keV at 1332 keV of 60Co peaks.

A Montecarlo based commercial code “EffCalcMC” was used for simulation of efficiency for different cylindrical geometries. The software is an extension of Nuclide Master Plus by Baltic Scientific [7] and the efficacy of the software are described elsewhere (Berlizov et al., 2006).[4] To generate the efficiency for a particular energy, this code uses detector parameters (dimension, window thickness of material, etc.), dimensions of container and sample height in the container. It also generates matrix specific efficiency such as for silica, and water by selecting from the material data base. Detector parameters were supplemented from the detector operation manual.

A major part of environmental samples like soil, sediment, sand, ore, etc., contain significant proportion of silica as their constituent. Self-absorption of photons of energy range between 50 and 200 keV is also significant in silica matrix.[5] Since the most commonly used cylindrical geometry for environmental sample at this laboratory is of 250 ml (diameter 7 cm, height 6.5 cm), therefore silica matrix and 250 ml cylindrical geometry were chosen for this study. [Figure 1]a shows the change in height of sample by changing the diameter of sample container of same volume (250 ml in this case) which leads to change in integral solid angle subtended by the sample to detector. [Figure 1]b shows the detector dimension and other parameters used for the calculation.
Figure 1: (a) S: Point source; A, B and C: Cylindrical containers of same volume and different diameter (here the volume of container is 250 ml and diameters of containers are 7, 9 and 11 cm); Ω: Solid angle tended by the point source; D: Window to detector distance (fixed detector parameter). (b) Detector parameters used for simulation of efficiency, all numbers are in mm (detector dimensions are provided by the manufacturer)

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The “EffCalcMC” Code was validated by comparing the experimental efficiency obtained using 250 ml plastic cylindrical (diameter 7 cm, height 6.5 cm) geometry for IAEA SRM RGU1 with code simulated efficiency for the same geometry and matrix. Using the code, efficiencies for 250 ml geometry cylindrical containers of diameters 5–16 cm were worked out. Diameters of 7 cm (existing geometry), 9 cm and 11 cm are compared for silica matrix efficiencies and are presented in this paper.


  Results and Discussion Top


The comparisons between the experimental efficiency and the code generated efficiency are shown in [Figure 2]. After validation of software, efficiencies were generated for energies 59.5, 63.3, 143.8 and 186.2 keV for different cylindrical geometries (250 ml) of diameters ranging from 5 to 16 cm.
Figure 2: Comparison between the efficiencies obtained for uranium series gamma energies (IAEA RGU-1) from experimental and EffCalcMC values

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Improvement in efficiency

In a gamma detector system, the detection efficiency depends on the number of gamma photons reaching the active detector volume relative to the photons emitted by source which is mainly a function of the solid angle subtended by the source at the detector. Another factor that affects the absolute detection efficiency is the self-attenuation of gamma photons within the sample matrix itself. When gamma radiation of intensity I0 passes through a sample medium of thickness t, the transmitted intensity I is given by the exponential expression, ; where μ1 is the linear attenuation coefficient (cm −1) which depends on the gamma energy, density and thickness of the medium. From the above expression, , we can get the relation between thickness of the sample medium and the transmitted gamma photons that will reach the detector as ln (I0/I)) α t. That means, a lower thickness of sample, results in better transmission of low gamma energies to reach the detector effective volume and hence an improved detection efficiency.[6]

[Figure 3] depicts the decrease in height for a 250 ml cylindrical container with different diameter and [Figure 4] shows the variation in efficiencies generated by EffCalcMC for 59.5, 63.3, 143.8 and 186.2 keV gamma energies for a sample of silica matrix for 250 ml cylindrical geometry with increasing diameters (5–16 cm). It has been observed that the efficiencies for all the selected energies were gradually increased up to 11 cm diameter and then started decreasing with higher values of diameters. This can be mainly attributed to two factors; the solid angle subtended by source to detector and the variation in self-attenuation due to change in thickness of sample. With increase in diameter, at first the solid angle increases, hence the number of photon reaching the detector. However later when the diameter of sample container becomes more than the diameter of detector window, photon loss occurs and it increases with further increase in the sample diameter. Again with decrease in thickness of sample, photon attenuation reduces and hence the graph shown in [Figure 4] is a combined effect of these two factors.
Figure 3: Diameter versus height for 250 ml cylindrical geometry container

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Figure 4: Variation of efficiencies for selected low gamma energies for different diameter of 250 ml cylindrical geometries in silica matrix by EffCalcMC

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Improvement in minimum detectable activity

An improvement in efficiency will result a lower minimum detection level, when all other parameters are fixed [Equation 1]. In [Table 1], a comparison has been made to show the effciencies at selected gamma energies for the 7 cm diameter (geometry being used at the laboratory) and geometries of 9 and 11 cm diameters. Efficiencies of 59.5, 63.3, 143.8, and 186.2 keV are increased by 10.37%, 16.32%, 20.53%, and 22.82%, respectively, for a change in diameter from 7 to 9 cm. For an increase in diameter from 9 to 11 cm, there is no significant improvement in efficiencies for energies 59.5 and 63.3 keV and the increase in efficiency for 143.8 and 186.2 keV are marginal. This is attributed to the fact that the change in height of sample medium for 250 ml containers of diameter 9 cm to that of 11 cm (1.30 cm) is less than for a change in height for container diameter of 7 –9 cm (2.57 cm) as can be seen in [Figure 3].
Table 1: Comparison of efficiencies for diameters 7, 9 and 11 cm and change in minimum detectable activity due to optimized diameter

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It is important that the sample container diameter should not be more than the detector diameter, to avoid positional errors while keeping the sample on the detector top. The diameter of the 50% R.E. HPGe detector system is approximately 9 cm. Hence, a 250 ml cylindrical geometry of 9 cm diameter is optimized for use. For this geometry, 9%–19% lower MDA can be achieved in the low gamma energy range of 50–200 keV.


  Conclusions Top


Achieving a lower detection limit for environmental sample analysis is highly desirable. In high resolution gamma spectrometry, the sample geometry and sample size play important role in analysis, as it directly affects the efficiency and hence the MDA. In this study, it was envisaged to improve the detection level of the system without compromising the sample size. A cylindrical geometry of 250 ml volume of 9 cm diameter is optimized to be used in place of the in-practice sample geometry of 7 cm diameter. The simulation study using the EffCalcMC software shows an improvement of 9%–19% in MDA for low energy gamma range of 50–200 keV. This can be attributed to the improved efficiencies resulted by the higher solid angle subtended by the optimized sample geometry and lower attenuation of gamma photons within the matrix.

Financial support and sponsorship

Nil.

Conflicts of interest

There are no conflicts of interest.

 
  References Top

1.
Jodlowski P. Self-absorption correction in gamma-ray spectrometry of environmental samples – An overview of methods and correction values obtained for the selected geometries. Nukleonika 2006;51:S21-5.  Back to cited text no. 1
    
2.
Gilmore GR. Practical Gamma Ray Spectrometry. 2nd ed. Warrington, UK: John Wiley & Sons, Ltd.; 2008.  Back to cited text no. 2
    
3.
AQCS. Intercomparison Runs Reference Materials. Analytical Quality Control Services. Vienna, Austria: IAEA; 1995.  Back to cited text no. 3
    
4.
Berlizov AN, Danilenko VN, Kazimirov AS, Solovyeva SL. True coincidence correction factor calculation for cascade gamma - Radiation based on statistical modeling with the use of evaluated nuclear data. At Energy 2006;100:382-8.  Back to cited text no. 4
    
5.
Galloway BR. Correction for sample self-absorption in activity determination by gamma spectrometry. Nucl Instrum Methods Phys Res 1991;A300:367-73.  Back to cited text no. 5
    
6.
Knoll GF. Radiation Detection and Measurements. 3rd ed. New York, USA: John Wiley & Sons, Ltd.; 2000.  Back to cited text no. 6
    
7.
Available from: http://www.lsrm.ru/en/products/detail.php?ELEMENT-CODE=nuclidemasterplus2-7. [ Last accessed on 2015 Nov 15].  Back to cited text no. 7
    


    Figures

  [Figure 1], [Figure 2], [Figure 3], [Figure 4]
 
 
    Tables

  [Table 1]



 

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