Source-detector Geometry Research Articles

Development of techniques for gantryless associated-particle imaging

To move fast-neutron radiography using the associated-particle imaging technique from the laboratory to the field, the development of new analysis techniques is required. In particular, the relative positions of the source and detectors need to be determined when they have been placed by hand, the normalization for a particular source–detector geometry needs to be determined without a measurement in the same geometry with no object present, and accurate image stitching is required when multiple detector positions are necessary to image an object. The present work describes methods that employ transmission neutron data to localize a fast-neutron imaging panel with respect to the neutron source, calculate a normalization for a given source–detector geometry, stitch images together, and describes the required system calibrations. The reported techniques enable in-field neutron radiography for cases in which the source–detector geometry is not well known a priori and where operational constraints preclude a normalization measurement.

A New Approach for NaI(Tl) Detector Efficiency Calibration Using 41Ar Radionuclide for Air Exhaust Systems

Argon-41 is an essential gaseous radionuclide that must be monitored in gaseous effluents from nuclear facilities. Therefore, a precise evaluation of 41Ar activity is highly desired. Gamma spectroscopy with a NaI(Tl) scintillation detector coupled with a multichannel analyzer (MCA) is one of the widely used techniques for the identification and activity measurements of radioisotopes. However, the efficiency calibration of these kinds of monitoring systems highly depends on the source-detector geometry, and a large amount of uncertainty may complicate the calibration. This paper presents the evaluation of the full peak efficiency of a 2 × 2-in. NaI(Tl) scintillation detector coupled with a stable MCA for a 41Ar source with 1293.5 keV energy in two different source-detector geometries, duct and Marinelli beaker, using the FLUKA code. A new experimental technique is considered to produce 41Ar in a controlled geometry, like a Marinelli beaker, through neutron irradiation of natural argon inside a cyclotron bunker. The simulation data were compared with the experimental results for Marinelli beaker geometry, and the ratio was evaluated as 0.99 ± 0.07. The ratio was considered a scaling factor for the final efficiency calibration of duct geometry.

Characterization of a bimodal multi-fibre optic clinical probe for in situ tissue diagnosis based on spectrally-and temporally-resolved autofluorescence

Introduction: Fibred-based optical spectroscopy is advantageous over imaging due to its sensitivity, practicality and precision, providing point of care diagnosis. The unique advantage is that the sampled volume is well defined by the source-detector geometry and that the functionality of multiple optical techniques can be incorporated into one probe so that more information is gained without extra bulkiness or cost, while also mitigating the limitations of each.Methods: This advantage is utilized here to address the limited accuracy in delineating brain tumors, in situ, by simultaneously characterizing tissue based on the spectral and lifetime properties of five endogenous fluorophores commonly present in brain tissue. A 5-meters multi fibre-optic probe custom-built for neurosurgery guidance with a sterilizable distal end is presented in this paper. It describes its technical features such as architecture, collection efficiency, sensitivity.Results and discussion: The developed probe have a temporal resolution of 34.6 ps and a temporal impulse response function of 29 ps. The spatial resolution was around 153 μm while the spectral resolution is 1 nm. All of these specifications are ensured under 1 second as acquisition time. The probe is validated using fluorescent dyes and tissue phantoms. Additionally, the fluorescence spectra from fresh and frozen animal tissue and from an epileptic patient are compared to that obtained from confocal fluorescence microscopy, while the lifetime values are compared to that reported in literature. Finally, the effect of pressure applied to the probe in contact with tissue is discussed.

Hybrid analytical method for calibrating a standard NaI(Tl) gamma-ray scintillation detector using a lateral hexagonal radioactive source

In various fields of radiation and nuclear physics, research and its applications help in the management and control of human activities, especially in the analysis of nuclear waste and measurement of low activity environmental samples. The current study uses a hybrid analytical technique and a 152Eu lateral hexagonal radioactive source to calibrate a standard cylindrical NaI(Tl) γ-ray scintillation detector. The technique is a combination of experimental measurements and theoretical calculations because for each environmental and/or radioactive waste sample, a separate reference calibration source equivalent to samples of different compositions is required, which needs additional funding. This method is suitable for any source–detector geometry. It can be used to calculate the geometric, total, and full-energy peak efficiency and the P/T ratio using the model probability of detecting γ-radiation photons, in particular from a lateral hexagonal radioactive source. The detection probability can be mainly influenced by the direction of the γ-quantum motion. This can help build more realistic and complex models than the typical point source case, where simple integration is used to numerically calculate the efficiency of the detection system. Using this method, the self-attenuation and coincidence-summing effect can be calculated for any gamma source matrices. The results obtained can show how useful this technique can be for calibrating complex gamma-ray detectors with any type of gamma-ray source.

Regression-based neural network for improving image reconstruction in diffuse optical tomography.

Diffuse optical tomography (DOT) is a non-invasive imaging technique utilizing multi-scattered light at visible and infrared wavelengths to detect anomalies in tissues. However, the DOT image reconstruction is based on solving the inverse problem, which requires massive calculations and time. In this article, for the first time, to the best of our knowledge, a simple, regression-based cascaded feed-forward deep learning neural network is derived to solve the inverse problem of DOT in compressed breast geometry. The predicted data is subsequently utilized to visualize the breast tissues and their anomalies. The dataset in this study is created using a Monte-Carlo algorithm, which simulates the light propagation in the compressed breast placed inside a parallel plate source-detector geometry (forward process). The simulated DL-DOT system's performance is evaluated using the Pearson correlation coefficient (R) and the Mean squared error (MSE) metrics. Although a comparatively smaller dataset (50 nos.) is used, our simulation results show that the developed feed-forward network algorithm to solve the inverse problem delivers an increment of ∼30% over the analytical solution approach, in terms of R. Furthermore, the proposed network's MSE outperforms that of the analytical solution's MSE by a large margin revealing the robustness of the network and the adaptability of the system for potential applications in medical settings.

Numerical Optimisation of a NIRS Device for Monitoring Tissue Oxygen Saturation.

The present work aims to develop a wearable, textile-integrated NIRS-based tissue oxygen saturation (StO2) monitor for alerting mobility-restricted individuals - such as paraplegics - of critical tissue oxygen de-saturation in the regions such as the sacrum and the ischial tuberosity; these regions are proven to be extremely susceptible to the development of pressure injuries (PI).Using a combination of numerical methods including finite element analysis, image reconstruction, stochastic gradient descent with momentum (SGDm) and genetic algorithms, a methodology was developed to define the optimal combination of wavelengths and source-detector geometry needed for measuring the StO2 in tissue up to depths of 3 cm. The sensor design was optimised to account for physiologically relevant adipose tissue thicknesses (ATT) between 1 mm and 5 mm. The approach assumes only a priori knowledge of the optical properties of each of the three tissue layers used in the model (skin, fat, muscle) based on the absorption and scattering coefficients of four chromophores (O2Hb, HHb, H2O and lipid).The results show that the selected wavelengths as well as the source-detector geometries and number of sources and detectors depend on ATT and the degree and volume of the hypoxic regions. As a result of a genetic algorithm used to combine the various optimised designs into a single sensor layout, a group of four wavelengths was chosen, coinciding with the four chromophores and agreeing very well with literature. The optimised number of source points and detector points and their geometry resulted in good reconstruction of the StO2 across a wide range of layer geometries.

Soil analysis using a 2″ NaI gamma detector

We describe an undergraduate physics laboratory experiment that uses a 2-inch diameter NaI gamma detector to measure natural radiation in soils. Students first calibrate the detector for the source-detector geometry, energy-dependent detection efficiency, and sample self-absorption. Then, the activity of the 238U and 232Th decay series, as well as the activity of 40K in the soil, are measured. The results are comparable in accuracy to those obtained using a high-resolution germanium gamma detector.

Footprint and height corrections for UAV-borne gamma-ray spectrometry studies

Advancements in the development of gamma-ray spectrometers (GRS) have led to small and lightweight spectrometers that can be used under unmanned aerial vehicles (UAVs). Airborne GRS measurements are used to determine radionuclide concentrations in the ground, among which the natural occurring radionuclides 40K, 238U, and 232Th. For successful applications of these GRS sensors, it is important that absolute values of concentrations can be measured. To extract these absolute radionuclide concentrations, airborne gamma-ray data has to be corrected for measurement height. However, the current analysis models are only valid for the height range of 50–250 m. The purpose of this study is to develop a procedure that correctly predicts the true radionuclide concentration in the ground when measuring in the UAV operating range of 0–40 m. An analytical model is developed to predict the radiation footprint as a function of height. This model is used as a tool to properly determine a source-detector geometry to be used in Monte-Carlo simulations of detector response at various elevations between 0 and 40 m. The analytical model predicts that the smallest achievable footprint at 10 m height lies between 22 and 91 m and between 40 and 140 m at 20 m height. By using Monte-Carlo simulations it is shown that the analytical model correctly predicts the reduction in full energy peak gamma-rays, but does not predict the Compton continuum of a spectrum as a function of height. Therefore, Monte-Carlo simulations should be used to predict the shape and intensity of gamma-ray spectra as a function of height. A finite set of Monte-Carlo simulations at intervals of 5 m were used for the analysis of GRS measurements at heights up to 35 m. The resulting radionuclide concentrations at every height agree with the radionuclide concentration measured on the ground.

Automated optical distance measurements for counting at a defined solid angle

Two optical distance measurement devices have been compared for accurately defining the source-detector geometry of alpha-particle counters. The first consists of a travelling microscope with unifocal lens, to which a linear distance gauge is attached. The focusing is done by human eye. The second is a modern 3D coordinate measurement machine, equipped with a CMOS colour camera as the sensor. Focusing is performed automatically by software. With both devices, distance measurements of various drop-deposited sources on glass and stainless steel disks have been performed. There is a good mutual agreement of results. Even though the new device can sample more reference points, the accuracy of the measurements has not improved drastically due to the imperfect flatness of the sources. The main advantages of the new system are its ease of use and the lower health risk involved due to the larger distance between the performer and the source.

Spectrum average cross section measurement of 183W (n, p)183Ta and 184W (n, p)184Ta reaction cross section in 252Cf(sf) neutron field

Neutron induced nuclear reactions are of prime importance for both fusion and fission nuclear reactor technology. Present work describes the first time measurement of spectrum average cross section of nuclear reactions 183W(n,p)183Ta and 184W(n,p)184Ta using 252Cf spontaneous fission neutron source. Standard neutron activation analysis (NAA) technique was used. The neutron spectra were calculated using Monte Carlo N Particle Code (MCNP). The effects of self-shielding and back scattering were taken into account by optimizing the detector modeling. These effects along with efficiency of detector were corrected for volume sample in the actual source-detector geometry. The measured data were compared with the previously measured data available in Exchange Format (EXFOR) data base and evaluated data using EMPIRE – 3.2.2.

Sensitivity analysis of high resolution gamma-ray detection for safeguards monitoring at natural uranium conversion facilities

Under the policies proposed by recent International Atomic Energy Agency (IAEA) circulars and policy papers, implementation of safeguards exists when any purified aqueous uranium solution or uranium oxides suitable for isotopic enrichment or fuel fabrication exists. Under IAEA Policy Paper 18, the starting point for nuclear material under safeguards was reinterpreted, suggesting that purified uranium compounds should be subject to safeguards procedures no later than the first point in the conversion process. In response to this technical need, a combination of simulation models and experimental measurements were employed in previous work to develop and validate gamma-ray nondestructive assay monitoring systems in a natural uranium conversion plant (NUCP). In particular, uranyl nitrate (UO2(NO3)2) solution exiting solvent extraction was identified as a key measurement point (KMP). Passive nondestructive assay techniques using high resolution gamma-ray spectroscopy were evaluated to determine their viability as a technical means for drawing safeguards conclusions at NUCPs, and if the IAEA detection requirements of 1 significant quantity (SQ) can be met in a timely manner. Building upon the aforementioned previous validation work on detector sensitivity to varying concentrations of uranyl nitrate via a series of dilution measurements, this work investigates detector response parameter sensitivities to gamma-ray signatures of uranyl nitrate. The full energy peak efficiency of a detection system is dependent upon the sample, geometry, absorption, and intrinsic efficiency parameters. Perturbation of these parameters translates into corresponding variations of the 185.7keV peak area of the 235U in uranyl nitrate. Such perturbations in the assayed signature impact the quality or versatility of the safeguards conclusions drawn. Given the potentially high throughput of uranyl nitrate in NUCPs, the ability to assay 1 SQ of material requires uncertainty «1%. Accounting for material self-shielding properties, pipe thickness, and source-detector orientation is instrumental in determining the robustness of gamma-ray detection in the process monitoring of uranyl nitrate in NUCPs. Monte Carlo models and ray-tracing models were employed to determine the sensitivity of the detected 185.7keV photon to self-shielding properties, pipe thickness, and source-detector geometry. Considering the implementation of the detection of 1 SQ, diversion of 1 SQ becomes essentially undetectable given the systematic uncertainty, in addition to considerations such as propagating uncertainties due to pipe offset/position, as well as minor variations in pipe thickness. Consequently, pipe thickness was the most sensitive variable in affecting full energy efficiency of the 185.7keV signature peak with up to 8% variation in efficiency for ±0.5mm changes in Schedule 40 304L stainless steel piping. Furthermore, computation of the attenuation correction factor of the uranyl nitrate solution [CF(AT) (i.e. εsample)] using Parker's method using with the approximation for the geometrical factor κ≈π/4 was validated through experimental, Monte Carlo and ray-tracing calculations for a uranyl nitrate filled transfer pipe segment. Quantifying sensitivity in detector position, as well as voiding effects due to bubbly flow or laminar flow with an air gap in the uranyl nitrate becomes increasingly important as considerations from (static) design-scale measurements translate into (dynamic) field operations tests.

Monte-Carlo simulation of a slot-scanning X-ray imaging system

We present a method for simulating slot-scanning X-ray imaging using the general-purpose Monte Carlo simulation package PENELOPE and penEasy Imaging. Different phantoms can be defined with the PENGEOM package, which defines bodies as combinations of volumes limited by quadric surfaces. The source-detector geometry, the position of the object, the collimator, the X-ray tube properties, the detector material and the pixel dimensions are defined. The output of the time-delay integration detector is simulated using sequential slot outputs derived from penEasy Imaging. The simulations are validated using tungsten and aluminium test objects, which are both simulated and imaged. The simulations are compared to the X-ray images using standard image quality metrics. The MTF, NPS and DQE curves show that the real and simulated X-ray images are comparable in terms of spatial resolution, noise and frequency information. The implementation can be modified to suit alterations in the system being simulated.

235U isotopic characterization of natural and enriched uranium materials by using multigroup analysis (MGA) method at a defined geometry using different absorbers and collimators

Abstract Characterization of nuclear materials is an important topic within the context of nuclear safeguards, homeland security and nuclear forensics. This paper deals with the performance of multigroup gamma-ray analysis (MGA) method using the X- and γ-rays in the 80-130 keV region and enrichment meter principle (EMP) based on the analysis of 185.7 keV peak for a certain geometry using different absorbers and collimators. The results from MGA and those of EMP are compared. In particular, the effect of aluminum/lead absorbers and lead collimator on the enrichment determination of 235U in natural and low enriched samples is investigated in a given source-detector geometry. The optimum diameter/height ratio for the Pb-collimator is found to be Dc/Hc = 1.4-1.6 in the chosen geometry. In order to simulate the container walls, ten different thicknesses of Al-absorbers of 141 to 840 mg·cm-2 and six different thicknesses of Pb-absorbers of 1120-7367 mg·cm-2 are interposed between sample and detector. The calibration coefficients (% enrichment/cps) are calculated for each geometry. The comparison of the MGA and EMP methods shows that the enrichment meter principle provides more accurate and precise results for 235U abundance than those of MGA method at the chosen geometrical conditions. The present results suggest that a two-step procedure should be used in analyses of uranium enrichment. Firstly MGA method can be applied in situ and then EMP method can be used at a defined geometry in laboratory.

Differential pathlength factor estimation for brain-like tissue from a single-layer Monte Carlo model.

A Monte Carlo simulation-based computational model has been developed for tracing the pathway of light within a single layer of tissue like bloodless human brain. A reflectance mode source-detector geometry is assumed to illuminate the tissue slab with an irradiation of a near infrared wavelength and to detect the re-emitted light intensity. Light is considered to be attenuated within tissue by scattering and absorption. The model has been used to predict the relationship of mean optical path of photons with variable source-detector geometry and thus, to determine a differential pathlength factor (DPF) of 5.66 for incident light of wavelength 810 nm.

The uncertainty of counting at a defined solid angle

Specific uncertainty components of counting at a defined solid angle are discussed. It is potentially an extremely accurate technique for primary standardisation of activity of alpha emitters and low-energy x-ray emitters. Owing to its reproducibility, it is very well suited for half-life measurements. Considered sources of uncertainty are 1) source–detector geometry, 2) solid-angle calculation, 3) energy loss and self-absorption, 4) scattering, 5) detection efficiency. Other sources of uncertainty, such as source weighing, counting, dead time and decay data are common to other standardisation methods. Statistical uncertainty propagation formulas are presented for the solid angle subtended by a circular detector to radioactive sources. Computer simulations were performed to investigate aspects of particle scattering.

Assessment of the uncertainty budget associated with 4πγ counting

The 4πγ-counting technique is recognized as a powerful primary method for the standardization of radionuclides decaying with abundant gamma emissions. Based on the use of a gamma detector in quasi 4π-geometry, a detection efficiency close to 100% and a low uncertainty can be achieved thanks to the summing effect of subsequent gamma transitions. Uncertainties have to be assigned to the realistic modelling of the source–detector geometry with respect to dimensions, density and material composition, the calculation of the total counting efficiency of the detector for the various emitted radiation, and the effect of possible flaws in the decay scheme of a radionuclide on the calculated total efficiency. Other uncertainty factors pertain to typical metrological sources of uncertainty, such as weighing, nuclear counting with pulse pile-up and system dead-time effects, impurity corrections, decay corrections, timing and frequency, etc. In order to ensure good metrological practices at NMIs, the uncertainties particular to the method are discussed.

On scattering effects for volume sources in low-energy photon spectrometry

In this study, different aspects of the Compton scattering inside volume sources are illustrated using experimental approach and Monte Carlo simulation. For the low-energy range (below 100keV) scattered events represents around 30% of the whole spectrum. Influence of the source-detector geometry is discussed. The scattering effects induce strong differences in spectrum shape for different geometrical conditions. This should influence efficiency transfer factors. A new approach is proposed, including the scattered events, to avoid complex peak area determination.

Fast neutron spectrometry using thick threshold detectors

This paper discusses the use of thick threshold activation detectors for the characterization of low intensity neutron fields. This technique has been applied to the determination of the spectral emission of a low activity (37 GBq) Am-Be source. The reaction rates induced by the neutrons emitted by this source in different thick metallic targets (Al, Si, Fe, In) have been measured in the following reactions: 27Al(n,p)27Mg, 27Al(n,α)24Na, 28Si(n,p)28Al, 56Fe(n,p)56Mn, 115In(n, n′)115mIn and 115In(n, γ)116mIn. Each measured reaction rate corresponding to a threshold detector response depends on the spectral emission of the source via a correcting factor. This factor, which takes into account the source detector geometry, the neutron attenuation and diffusion by the detectors, has been determined by Monte Carlo simulation using MCNP5 code. The spectral emission of the neutron source has been generated from the response matrix of the threshold detectors by using different neutron spectrum unfolding methods (Stayn'l, Gravel and Maxed). A fairly good agreement with the assumed ISO spectrum has been achieved.

TV-Based image reconstruction of multiple objects in a fixed source-detector geometry

In this paper, we study how to reconstruct multiple objects tomographically from their overlapped x-ray projections. We propose a new rotation-translation scanning mode for a fixed source-detector geometry, in which the objects are individually translated periodically while they are rotated during the data acquisition. The reconstruction scheme performs SART reconstruction and TV minimization alternatingly. The scanning parameters are evaluated in a series of numerical experiments. Promising results are obtained in the cases of two and four objects respectively. The major advantage of our approach is that the scanning time can be reduced by about 46% (for two objects) and 67% (for four objects) respectively, improving the imaging facility throughput significantly.

A simple approach to determine the radioactivities of nuclides with 200–2000 keV γ ray(s) in the absence of efficiency calibration standard

By combing 4π liquid scintillation (LS) and γ spectrometry, we proposed a new approach, which is based on the linearity of the energy and the efficiency of coaxial HPGe detectors in logarithmic coordinates, to determine the radioactivities of nuclides with 200–2000 keV γ ray(s). In this approach, the multi-gamma-ray nuclide 152Eu has been applied to obtain the slope of energy-efficiency line. Meanwhile, the intercept of the energy-efficiency line is obtained by determining the 661.657 keV γ ray radiation of 137Cs whose radioactivity is determined by 4π LS. For the fixed source-detector geometry of this study, the energy-efficiency line is: lgɛγ = −0.9503lgEγ + 0.5224. This simple approach is promising to be used to determine the radioactivities of all the nuclides which emit γ ray(s) with the energy (energies) between 200 and 2000 keV in the absence of efficiency calibration standard. The relative standard deviation of the approach is about 5.7%.