Counting Geometry Research Articles

Development of a 3D human body library based on polygonal mesh surface for whole body counter set-up calibration

A Whole Body Counter (WBC) is a common facility to routinely assess the internal contamination of exposed workers and especially in the case of radiation release accidents. The calibration of the counting device is usually done by using anthropomorphic physical phantoms representing the human body. Due to such a challenge of having representative physical phantoms a virtual calibration has been introduced. The use of computational phantoms and Monte Carlo methods to simulate radiation transport have been demonstrated to be a worthy alternative, not only as a complementary calibration method but also as a powerful tool for the design and optimization of counting geometries. In this study we introduce a methodology for building a 3D human body library that spans variation in both subject shape and size. Open source codes as MakeHuman and Blender software packages have been used for the creation and modelling of 3D humanoid characters based on polygonal mesh surfaces. Also, a home-made software was developed whose main goal is to convert the binary 3D voxel grid into a MCNPX input file. This paper summarizes (1) the development of the MaMP and FeMP - Male and Female Mesh Phantoms, the two first mesh phantoms generated from a series of 3D human bodies library created at SCK-CEN and (2) presents a comparison between the counting efficiency results obtained from the simulations of both mesh phantoms using MCNPX code.

Certification of Total Arsenic in Blood and Urine Standard Reference Materials by Radiochemical Neutron Activation Analysis and Inductively Coupled Plasma - Mass Spectrometry.

A newly developed procedure for determination of arsenic by radiochemical neutron activation analysis (RNAA) was used to measure arsenic at four levels in SRM 955c Toxic Elements in Caprine Blood and at two levels in SRM 2668 Toxic Elements in Frozen Human Urine for the purpose of providing mass concentration values for certification. Samples were freeze-dried prior to analysis followed by neutron irradiation for 3 h at a fluence rate of 1×1014cm-2s-1. After sample dissolution in perchloric and nitric acids, arsenic was separated from the matrix by extraction into zinc diethyldithiocarbamate in chloroform, and 76As quantified by gamma-ray spectroscopy. Differences in chemical yield and counting geometry between samples and standards were monitored by measuring the count rate of a 77As tracer added before sample dissolution. RNAA results were combined with inductively coupled plasma - mass spectrometry (ICP-MS) values from NIST and collaborating laboratories to provide certified values of (10.81 ± 0.54) μg/kg and (213.1 ± 0.73) μg/kg for SRM 2668 Levels I and II, and certified values of (21.66 ± 0.73) μg/kg, (52.7 ± 1.1) μg/kg, and (78.8 ± 4.9) μg/kg for SRM 955c Levels 2, 3, and 4 respectively. Because of discrepancies between values obtained by different methods for SRM 955c Level 1, an information value of < 5 μg/kg was assigned for this material.

Measurement and calibration of metal and non-metal wastes produced from decommissioning

This report described a radioactive waste reference drum which was established with large-area sources and metal slices. This reference drum could be applied in calibration or testing of drum counting systems having 4π counting geometry and being made with plastic scintillators. This metal reference drum has the advantages of easy operation, low natural background and it also has agreeable measurement efficiency calibration curves for the drum counting system as the non-metal reference drum studied previously. On the other hand, this study explored the counting efficiency variations of the drum counting system by simulations of the metal reference drum being filled with wastes up to different heights within the drum. With the exploration, it is feasible to correct the measurement errors caused by different quantities of waste filling.

A detailed procedure to simulate an HPGe detector with MCNP5

Due to its high resolution, HPGe detectors are widely used for analysis of gamma emitters radioisotopes. The determination of the response curves for this type of detector is not easy and demands a large number of gamma emitters in order to account for the energy range of interest. For volumetric geometries, a standard solution of a mix of radionuclides is commonly used, but requires one standard solution for each counting geometry of interest. The Monte Carlo method can be used to determine the detector's response curves, making it easier and cheaper. This work presents a detailed description of the procedure to simulate and calibrate co-axial HPGe detectors. It also presents a complete input file for the MCNP5 computer code. The comparison of the simulated and the experimental data showed very good agreement and the discrepancies are mainly due to the uncorrected peak sum effect of the experimental data.

Monte Carlo Simulation of In Vivo Measurement of the Most Suitable Knee Position for the Optimal Measurement of Activity

A new computational model has been developed using the Monte Carlo (MC) technique to simulate in vivo measurements with the objective of understanding the most precise measurement location with respect to quantifying the activity of Am in the bones. To benchmark the model, in vivo measurements were performed on the U.S. Transuranium and Uranium Registries (USTUR) case 0846 leg. Front and lateral measurements of the knee of the USTUR case 0846 leg in a bent position and the same measurements with the leg in a straight position using a HP(Ge) detector were completed. Experimental results concluded that the front measurement of the knee in a bent leg position gave the highest count rate, which is an indication of optimal detection efficiency. Therefore, this geometry and knee-detector position were considered as the most appropriate position for knee monitoring. A computational model using MCNPX version 2.6.0 was used to simulate the experimental measurements by using a leg voxel phantom. The mean value and standard deviation (SD) of peak efficiency due to an isotropic 59.5-keV photon from Am were calculated in four different counting geometries. An extra sum of squares F-test was performed on the mean values of the simulated detection efficiencies. The p-value obtained from this statistical test indicates that the differences among the mean values for different counting geometries were significant. These results suggest that the front measurement of a knee in a bent leg position is the optimal counting geometry for in vivo measurement of Am deposited in the bones. The computational model was validated through comparison of the measured and simulated detection efficiencies. It was observed that there is no difference at the 0.1 significant levels between the simulated and measured detection efficiencies in assessment of Am within the bones.

Half-lives of 221Fr, 217At, 213Bi, 213Po and 209Pb from the 225Ac decay series

The half-lives of 221Fr, 217At, 213Bi, 213Po, and 209Pb were measured by means of an ion-implanted planar Si detector for alpha and beta particles emitted from weak 225Ac sources or from recoil sources, which were placed in a quasi-2π counting geometry. Recoil sources were prepared by collecting atoms from an open 225Ac source onto a glass substrate. The 221Fr and 213Bi half-lives were determined by following the alpha particle emission rate of recoil sources as a function of time. Similarly, the 209Pb half-life was determined from the beta particle count rate. The shorter half-lives of 217At and 213Po were deduced from delayed coincidence measurements on weak 225Ac sources using digital data acquisition in list mode. The resulting values: T1/2(221Fr)=4.806 (6) min, T1/2(217At)=32.8 (3)ms, T1/2(213Bi)=45.62 (6)min, T1/2(213Po)=3.708 (8)μs, and T1/2(209Pb)=3.232 (5)h were in agreement only with the best literature data.

Evaluation of a Commercial Software Package’s Ability to Calibrate an In Vivo Counter for Emergency Response

A commercial detector calibration package has been assessed for its use to calibrate the Human Monitoring Laboratory's Portable Whole Body Counter that is used for emergency response. The advantage of such a calibration software is that calibrations can be derived very quickly once the model has been designed. The commercial package's predictions were compared to experimental point source data and to predictions from Monte Carlo simulations. It was found that the software adequately predicted the counting efficiencies of a point source geometry to values derived from Monte Carlo simulations and experimental work. Both the standing and seated counting geometries agreed sufficiently well that the commercial package could be used in the field.

Use of sum-peak and coincidence counting methods for activity standardization of 22Na

A solution containing the positron emitter 22Na has been absolutely standardized using the 4πβ–γ coincidence counting method and the sum-peak spectrometry counting method. In the 4πβ–γ coincidence method two ways for the activity concentration measurements were used: gating on the 1275keV photopeak and on the 1786keV sum-peak where the knowledge of the β+-branching ratio is required. In the sum-peak method the measurements were carried out using three experimental arrangements: the first composed by a well type 5in.×5in. NaI(Tl) scintillation crystal, the second by a 3in.×3in. NaI(Tl) scintillation crystal placed on the top of the first, resulting in a 4π counting geometry and the third arrangement is a high purity coaxial germanium detector. The results that are obtained by these two methods are compatible within the standard uncertainty values with a coverage factor of k=2 (∼95% of the confidence level). This means that the sum-peak counting with its more simple experimental setup than the complex coincidence 4πβ–γ counting system gives consistent results for the activity standardization of 22Na with smaller uncertainties. Besides, the time period involved to attain the result of the standardization was quite shorter than the coincidence measurements used in this work.

HPGe well detector calibration procedure by MCNP5 Monte Carlo computer code

The determination of response curves for HPGe detector demand a large number of gamma sources in order to account for the energy range of interest for a given counting geometry. HPGe well detectors present an additional difficulty, related to the large sum effect for sources placed inside the well, either limiting the useful sources or large analytical corrections, which might imply in large uncertainties. The Monte Carlo method can be used to determine a detector’s response curves which are difficult to obtain experimentally. The MCNP5 computer code provides means to simulate gamma ray detectors and has been used for this work for the 50–2000keV energy range. The detector’s sensitive volume was determined by comparison of simulated and measured spectra due to two point sources, 241Am and 137Cs. The thickness of the dead layer+transition layer was determined for all crystal’s surfaces, and the response curves for two volumetric sources geometries, one placed on top of the detector and a smaller one placed inside the well, were then obtained by simulation and compared with the experimental results, in order to verify and validate the detector’s simulation. Both simulated and experimental response curves were in very good agreement.

Accounting for detector crystal edge rounding in gamma-efficiency calculations theoretical elaboration and application in ANGLE software

In absolute and semi-empirical calculations of full gamma-energy peak efficiencies (ep), geometrical/compositional data characterizing the detector should be known in much detail. Among these, detector crystal edge rounding (bulletization), if neglected, may lead to large systematic errors, especially for low gamma-energies and close counting geometries. The errors show quadratic dependence on the extent of bulletization (bulletization radius). Mathematical/analytical solution to the problem - not reported so far - is elaborated in the present work. Relevant mathematical formulae are derived for a number of counting arrangements most frequently encountered in gamma-spectrometry practice (point, disc, cylinder, and Marinelli sources). These are subsequently programmed for numerical calculations and are now part of commercially available ANGLE software. Extensive calculation testing is performed for HPGe (both p- and n-type) and LEPD detectors (several sizes each), with various sources (point, disc, cylinder, Marinelli) and counting geometries (0-20 cm source-to-detector distance). Energy range considered was 10-3000 keV. To elucidate the significance of the issue, an error propagation study was conducted: results with bulletization taken into account are compared to those when bulletization was neglected. Corresponding errors are tabulated in an extensive Excel file. The file comprises about 152 000 error calculation results which are available for download; a few characteristic ones are selected for presenting in the paper. ANGLE proved handy (in programming) and fast (in calculations) when accomplishing this task. The data convincingly illustrate the impact of detector bulletization on gamma-efficiency and thus the need to account for. Even only slight bulletization (1-2 mm bulletization radius) is not negligible in many realistic counting situations. Reader/analyst can (1) compare his/her counting situation with these data so as to get the first impression of the problem and (2) use the mathematical model presented and/or ANGLE software to address the issue.

MCNPX alpha particle dose estimate to the skin tissue from a low-enriched uranium fuel fragment

Three alpha volume sources (low-enriched uranium-U(3)Si) were analysed using Monte Carlo modelling in order to calculate the dose delivered to the dermis from a small embedded fuel fragment (sliver). Three shapes were analysed using MCNPX 2.6.0 code: sphere, cylinder and parallelepiped. Essentially, two kinds of runs were performed: count rate run and dosimetry run. The two results were combined to estimate dose coefficients that can be used for alpha dose assessments in the field. The two results were obtained for the 1 and 0 cm counting geometries. These results are very stable and show that the actual dose delivered to the skin per unit count rate for the recovered particle is independent of the shape of the volume alpha source.

Production of &lt;sup&gt;128&lt;/sup&gt;I Atoms through &lt;sup&gt;127&lt;/sup&gt;I (n,&amp;gamma;) &lt;sup&gt;128&lt;/sup&gt;I Reactions in some Solid Iodocompounds using an (&lt;sup&gt;241&lt;/sup&gt;Am-Be) Neutron Source

Present work reports the production of 128I atoms through 127I (n,g) 128I reactions in six iodine containing compounds (viz. NaI, NaIO3, NaIO4, KI, KIO3 and KIO4) in solid state. An 241Am-Be neutron source has been used for the production of 128 I atoms. The radioactivities produced have been measured by a thin end-window G.M. Counter. The irradiation time and counting geometry were constant throughout the present sets of experiments. Results obtained have been analyzed through time-decay analyses. From the decay curves, the half life of 128I has been estimated ~25 minutes. The average zero time activities of 128I from three independent measurements in each of NaI, NaIO3.H2O, NaIO4, KI, KIO3 and KIO4 target, have been found 116 ± 2, 92 ± 4, 88 ± 6, 102 ± 6, 98 ± 1 and 93 ± 5 cpm respectively. It is interesting to note that the activities show a decreasing trend while passing through either from NaI to NaIO4 or from KI to KIO4. Results are attributed to the quantity of 128I produced in a specified volume element and, self absorption and self scattering of radiations within the target compounds studied. Keywords: Iodocompounds; 128I; 241Am-Be neutron source; Chemical effects of nuclear transformation.© 2011 JSR Publications. ISSN: 2070-0237 (Print); 2070-0245 (Online). All rights reserved.doi:10.3329/jsr.v3i2.6385 J. Sci. Res. 3 (2), 357-366 (2011)

Microchannel plate special nuclear materials sensor

Abstract Nova Scientific Inc., is developing for the Domestic Nuclear Detection Office (DNDO SBIR #HSHQDC-08-C-00190), a solid-state, high-efficiency neutron detection alternative to 3He gas tubes, using neutron-sensitive microchannel plates (MCPs) containing 10B and/or Gd. This work directly supports DNDO development of technologies designed to detect and interdict nuclear weapons or illicit nuclear materials. Neutron-sensitized MCPs have been shown theoretically and more recently experimentally, to be capable of thermal neutron detection efficiencies equivalent to 3He gas tubes. Although typical solid-state neutron detectors typically have an intrinsic gamma sensitivity orders of magnitude higher than that of 3He gas detectors, we dramatically reduce gamma sensitivity by combining a novel electronic coincidence rejection scheme, employing a separate but enveloping gamma scintillator. This has already resulted in a measured gamma rejection ratio equal to a small 3He tube, without in principle sacrificing neutron detection efficiency. Ongoing improvements to the MCP performance as well as the coincidence counting geometry will be described. Repeated testing and validation with a 252Cf source has been underway throughout the Phase II SBIR program, with ongoing comparisons to a small commercial 3He gas tube. Finally, further component improvements and efforts toward integration maturity are underway, with the goal of establishing functional prototypes for SNM field testing.

Evaluation of Radiochemical Neutron Activation Analysis Methods for Determination of Arsenic in Biological Materials

Radiochemical neutron activation analysis (RNAA) with retention on hydrated manganese dioxide (HMD) has played a key role in the certification of As in biological materials at NIST. Although this method provides very high and reproducible yields and detection limits at low microgram/kilogram levels, counting geometry uncertainties may arise from unequal distribution of As in the HMD, and arsenic detection limits may not be optimal due to significant retention of other elements. An alternate RNAA procedure with separation of arsenic by solvent extraction has been investigated. After digestion of samples in nitric and perchloric acids, As(III) is extracted from 2 M sulfuric acid solution into a solution of zinc diethyldithiocarbamate in chloroform. Counting of (76)As allows quantitation of arsenic. Addition of an (77)As tracer solution prior to dissolution allows correction for chemical yield and counting geometries, further improving reproducibility. The HMD and solvent extraction procedures for arsenic were compared through analysis of SRMs 1577c (bovine liver), 1547 (peach leaves), and 1575a (pine needles). Both methods gave As results in agreement with certified values with comparable reproducibility. However, the solvent extraction method yields a factor of 3 improvement in detection limits and is less time-consuming than the HMD method. The new method shows great promise for use in As certification in reference materials.

Female workers and in vivo lung monitoring: a simple model for morphological dependence of counting efficiency curves

This paper addresses the question of the morphological dependence of counting efficiency curves for in vivo lung monitoring of workers, with a particular focus on the case of female workers for whom different chest girth and cup size are considered. A library of 24 female torsos, with chest girth varying from 85 to 120 and cup size from A to F, was constructed using mesh and NURBS formats. The anatomical realism and usefulness of these models for simulating in vivo counting measurements are illustrated and simulations are reported for a typical 4-germanium (Ge) counting system. A simple analytic formula describing the relation between efficiency curves obtained for each female phantom is given. This formula uses the mass attenuation coefficient for adipose tissue and two parameters which are dependant on lung volume and breast weight. The model is tested against Monte Carlo simulated data, experimental data obtained with the Livermore phantom and published data. The model correctly describes the efficiency curve and, since the parameters depend on the counting geometry, it is shown how to estimate them from experimental measurements.

Rapid, non-destructive carbon analysis of forest soils using neutron-induced gamma-ray spectroscopy

Forest soils are pivotal to understanding global carbon (C) cycling and evaluating policies for mitigating global change. However, they are very difficult to monitor because of the heterogeneity of soil characteristics, the difficulty of representative sampling, and the slow time scale of response to environmental change. Here we demonstrate that use of gamma-ray spectroscopy facilitates in situ non-destructive analysis of C and other elements in forest soils. In this approach the element-specific gamma-rays are induced by fast and thermal neutrons interacting with the nuclei of the elements present in the soil. Background gamma-rays emanating from naturally occurring radionuclides in the forest are recorded as well. We applied this approach in a mature northern hardwood forest on glacial till soils at the Bartlett Experimental Forest in New Hampshire, USA. The inelastic neutron scattering (INS) system yielded strong signals in gamma-ray counts/h, from C and other elements present in the soil matrix that included silicon, oxygen, hydrogen, iron, aluminum, manganese and potassium. The INS sensitivity for carbon was 20.656 counts h −1 kg −1 C m −2 based on current net C gamma-ray counts and the data for the O horizon and mineral soil to a depth of 30 cm obtained from a nearby quantitative soil pit (7.35 kg C m −2). We estimate the minimum detectable change to be ∼0.34 kg C m −2, which is ∼5% of the current soil C content, and the minimum detectable limit to be ∼0.23 kg C m −2. Eight % reproducibility from 11 measurements was limited, in part, by the large variability in the system counting geometry due to the uneven forest microtopography. The INS approach has the potential to revolutionize belowground monitoring of C and other elements, because the possibility of detecting a 5% change in forest soils has not been possible with destructive sampling methods.

Determination of radioactivity concentration in the primary coolant channel of a Pressurized Heavy Water Reactor

CdZnTe detector was used to monitor the radioactivity at several locations of primary heat transport (PHT) system of the Tarapur Atomic Power Station-III & IV reactor during shut down as well as operating condition of the reactor. Dose measurements at the same locations were also carried out using a teletector. MCNP code was used to simulate the detector efficiency for the required counting geometry. The efficiencies were then used to convert the experimental count rates (cps) to corresponding activity concentration (Bq/mL) of different radionuclides in coolant water. The total dose rates in these locations were also calculated by summing up the simulated dose rates of individual γ ray energy. The simulated and the measured total dose rates were found to be in reasonably good agreement.

True coincidence-summing corrections for the coincident γ-rays measured with coplanar grid CdZnTe detectors

In this study, true coincidence-summing (TCS) correction factors have been measured for the sources (22)Na, (60)Co, (133)Ba and (152)Eu by use of three large volume coplanar grid CdZnTe (acronym: CZT) detectors. In case of a close-in detection geometry, two different TCS calculation algorithms were used to compute the required TCS correction factors. Both of the algorithms are based on the measured total-to-peak (TTP) ratio and full-energy peak (FEP) efficiency values that were obtained using almost "single" energy and coincidence-free nuclides. The results for TCS correction factors obtained by two different algorithms were agreeable to each other. The obtained TCS factors were ranged from about 7% to 30.5% in a 2250 mm(3) CZT detector when a close counting geometry was used. For other two detectors with a volume of 1000 and 1687.5mm(3), the resulted TCS correction factors were relatively smaller and varied between about 0.1% and 20% at the close counting geometry condition. Therefore, the results indicate that there is a need for the estimation of TCS corrections in CZT detectors, especially when their crystal volumes are greater than 1cm(3) and these detectors are used in the case of a close-in detection geometry.

The 124Sb activity standardization by gamma spectrometry for medical applications

This work describes a metrological activity determination of 124Sb, which can be used as radiotracer, applying gamma spectrometry methods with hyper pure germanium detector and efficiency curves. This isotope with good activity and high radionuclidic purity is employed in the form of meglumine antimoniate (Glucantime) or sodium stibogluconate (Pentostam) to treat leishmaniasis. 124Sb is also applied in animal organ distribution studies to solve some questions in pharmacology. 124Sb decays by β-emission and it produces several photons (X and gamma rays) with energy varying from 27 to 2700 keV. Efficiency curves to measure point 124Sb solid sources were obtained from a 166mHo standard that is a multi-gamma reference source. These curves depend on radiation energy, sample geometry, photon attenuation, dead time and sample-detector position. Results for activity determination of 124Sb samples using efficiency curves and a high purity coaxial germanium detector were consistent in different counting geometries. Also uncertainties of about 2% ( k=2) were obtained.

Use of photographic images to construct voxel phantoms for use in whole-body counting

Quantification of radioactivity in the body by in vivo bioassay uses counting efficiencies obtained from calibration from a phantom. Usually a standardised BOMAB (Bottle Manikin Absorption) phantom is employed for whole-body counting. The physical size of workers being counted, however, may differ from the calibration phantom, and can be a source of significant errors in dose estimates. A methodology was developed applying subject-specific efficiency data determined by Monte Carlo simulation based on a voxel phantom that was constructed from photographic images of the subject. This approach was demonstrated using a BOMAB phantom. The measured and calculated efficiencies agreed well, with maximum deviation of 30 % at 1.836 MeV (Y-88 gamma-rays). The expected counting efficiencies for an obese volunteer appear higher compared with a BOMAB phantom. This is caused by a closer distance between the detector and the body surface. The fast construction technique of voxel phantoms will contribute to a reduction in uncertainty caused by variations in the counting geometry.