Attenuation correction of polychromatic L-shell X-ray fluorescence computed tomography imaging

Polychromatic L-shell X-ray fluorescence computed tomography is a promising imaging technique to explore the element distribution of the sample. However, the image quality of large samples would be decreased dramatically due to the present of the self-absorption. In this investigation, an attenuation correction algorithm based on theories related to X-ray fluorescence was proposed. With this method, the attenuation coefficients at the incident energies were expressed as a function of known X-ray energies and unknown platinum concentrations. Then the attenuation coefficients were calculated based on the theories and added in the contribution value of the pixel in Maximum Likelihood Expectation Maximization reconstruction method. The element distributions can be got through continuous iterations. Finally, the feasibility of this method was tested by Monte Carlo simulation. The results indicated that, for a 4-mm diameter phantom containing an object of 2 mm in diameter with 0.10 wt% platinum solutions, the CNR of reconstructed images increased from 53.5 to 203.8 after attenuation correction while the relative error decreases from 97% to 2%.     

Self-absorption correction for low energy gamma rays: Application to 210Pb determination in marine sediments

A series of nondestructive, low energy gamma-ray spectrometric measurements by means of intrinsic Ge detectors was carried out for the activity determinations of several naturally occurring radionuclides in marine sediments. The system was calibrated for efficiency using a radioactive sample obtained by combining standard radioactive solutions with a suitable matrix. This procedure is strictly only correct for the analysis of samples of similar composition with the same mass absorption coefficients at the investigated energies. Nevertheless, the efficiency calibration can be extended to different matrix samples if the variation in self-absorption is accurately taken into account.A method for calculating the self-absorption correction at any energy of interest is proposed and verified for the determination of ²¹⁰Pb activity at an energy of 46.5 keV. The direct gamma method and the conventional, accurate, but tedious, alpha technique via ²¹⁰Po, which can be alternatively applied to the determination of ²¹⁰Pb, agreed mostly within 8%.     

高纯锗探测器探测效率研究

利用系列标准γ射线源对高纯锗探测器的探测效率进行了各种测量,与蒙特卡罗计算程序相结合,对于高纯锗探测效率进行了分析和讨论.计算效率与测量效率在4%以内吻合.在一定探测距离条件下面源与点源的探测效率在1%以内吻合,而且面源的自吸收可以用平行束在材料中的自吸收来计算;当面源靠近探测器时,由于γ射线的倾斜入射,这种方法就不适用了,需要用蒙特卡罗方法进行自吸收较正.     

A method to determine the gamma-ray linear attenuation coefficient

In this work, gamma-ray linear attenuation coefficients of the materials such as Pb, Fe, Cu and Al were determined by means of traditional energy method and the developed timing detection technique. 1.33 MeV-energy radiation emitted from ⁶⁰Co radioisotope was used in the experiments. The gamma-ray linear attenuation coefficient values of the materials were calculated by using Xcom code, and the experimental results were compared with the theoretical ones.     

Determination of within-rod caesium and europium isotopic distributions in high burnup fuel rods through computerised gamma-ray emission tomography

Computerised gamma-ray emission tomography has been applied to single PWR UO₂ fuel rods, with pellet averaged burnups of 52, 71, 91 and 126 GWd/t respectively, for the determination of ¹³⁴Cs, ¹³⁷Cs and ¹⁵⁴Eu internal radial distributions. State-of-the-art image reconstruction techniques, analytical and iterative, have been applied, evaluated and compared using test phantoms first and, in a second step, the actual measured data. Further, linear attenuation maps, previously derived on the same samples by means of gamma-ray transmission tomography, have been used to correct for density inhomogeneities. The final results have indicated large central depressions in the caesium distributions, but of varying extent from sample to sample. Particularly interesting is the case of the 126 GWd/t sample, showing a very deep central depression (periphery-to-centre ratios of ∼2.5 for ¹³⁷Cs and ∼3 for ¹³⁴Cs). In addition, a difference in the relative activity distributions of ¹³⁷Cs and ¹³⁴Cs has been observed for all the samples. In contrast, the europium shows an almost flat distribution.     

Measurements and statistical model calculations of activation cross-sections for 26Mg(n,α)23Ne reaction between 13.6 and 14.9 MeV neutron energies

Activation cross-sections were measured at neutron energies from 13.6 to 14.9 MeV for the reaction ²⁶Mg(n,α)²³Ne. The production of relatively short-lived activity and the spectra accumulation have been carried out by the cyclic activation method. Corrections were made for the effects of gamma-ray attenuation, random coincidence summing (pulse pile-up), dead time, and scattered low energy neutron contribution. Statistical model calculations for which the pre-equilibrium emission effects are taken into consideration were also performed. Results were compared with the previous investigations.     

Measurements of activation cross-sections for the F(n, α)N reaction for neutrons with energies between 13 and 15 MeV

In this study, activation cross-sections were measured for the ¹⁹F(n, α)¹⁶N reaction at six different neutron energies from 13.5 and 14.9 MeV. The fast neutrons were produced via the ³H(d,n)⁴He reaction on SAMES T-400 neutron generator. The cyclic activation technique was used. Induced gamma activities were measured by a high-resolution gamma-ray spectrometer with high-purity germanium (HpGe) detector. Measurements were corrected for gamma-ray attenuations, random coincidence (pile-up), dead time and fluctuation of neutron flux. Results were compared with the previous works.     

Calculations of the transport of neutrons and secondary gamma rays through concrete for incident neutrons in the energy range 15 to 75 MeV

Earlier work by Alsmiller et al. considered coupled neutron and secondary-gamma-ray transport through a thick shield of silicon dioxide with 5% water by weight for neutron sources with energies of 50, 100, 200, 300 and 400 MeV. In that work, the approximation was made that gamma rays were produced only by neutron capture. In the present work, coupled neutron and secondary-gamma-ray transport through a thick shield of concrete for neutron sources with energies of 15, 25 and 75 MeV is considered. In this study, gamma-ray production for all interactions involving neutrons with energies up to 15 MeV was included; i.e., the approximation made here is that gamma-ray production can be neglected for interactions by neutrons with energies > 15 MeV.For incident neutron energies of 15, 25, 50, and 75 MeV, results of total and gamma-ray dose equivalents are given as a function of depth into the slab. For the 50- and 75-MeV incident neutron energies, the gamma-ray dose equivalent was found to be no more than 5% of the total dose equivalent at all depths considered ( 1500 g/cm²). For the 15- and 25-MeV incident neutron energies, however, the gamma-ray dose equivalent dominates at greater depths into the slab. A conservative estimate of the effect of including gamma rays produced in interactions with neutrons of energies > 15 MeV indicates that the calculated total dose equivalent would increase by no more than 5%.     

Chemical corrosion and gamma-ray attenuation properties of Zr and Ti containing lead silicate glasses

Lead silicate glasses (LSGs) have high gamma-ray attenuation but low chemical durability properties. In this work, LSGs with (55.5-68.5 wt%) PbO content containing ZrO₂ and TiO₂ additions were produced. The chemical corrosion of various produced LSGs in 0.5 N HNO₃ aqueous solution and determination of their gamma-ray attenuation coefficients for ⁶⁰Co and ¹³⁷Cs sources were investigated. The weight loss measurements, the SEM micrographs, the EDS analysis of the sample surfaces and the ICP analysis of solution were used to characterize the dissolution process. The effects of PbO content, ZrO₂ and TiO₂ additives on chemical corrosion, and also the effect of PbO on gamma-ray attenuation coefficient, glass transition temperature (Tg), and density of LSG glasses were determined. The results showed that by increasing the lead content of glass the gamma-ray attenuation coefficient, chemical corrosion and density were increased, but the Tg decreased. One of the samples with PbO contents of 65.4 wt% and SiO₂ content of 26.9 wt% showed a very low chemical corrosion behavior and good gamma-ray absorption property.     

A comprehensive study for mass attenuation coefficients of different parts of the human body through Monte Carlo methods

The gamma-ray mass attenuation coefficients of blood,bone,lung,eye lens,adipose,tissue,muscle,brain and skin were calculated at different energies(60,80,150,400,500,600,1000,1250,1500,and 2000 keV) by various theoretical methods such as FLUKA,GEANT4 Monte Carlo(MC) methods and XCOM program in this work.Calculated coefficients were also compared with the National Institute of Standards and Technology(NIST) values.Obtained results were highly in accordance with each other and NIST values.Our results showed that FLUKA was quite convenient in comparison to GEANT4 in the calculation of the mass attenuation coefficients of the used human body samples for low-energy photons(60,80,and 150 keV) when compared with the NIST values.     

Gamma ray and neutron shielding properties of some concrete materials

Shielding of gamma-rays and neutrons by 12 concrete samples with and without mineral additives has been studied. The total mass attenuation and linear attenuation coefficients, half-value thicknesses, effective atomic numbers, effective electron densities and atomic cross-sections at photons energies of 59.5 and 661 keV have been measured and calculated. The measured and calculated values were compared and a reasonable agreement has been observed. Also the recorded values showed a change with energy and composition of the concrete samples. In addition, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ_R, cm⁻¹) concept. The WinXCom and NXcom programs were employed to calculate the attenuation coefficients of gamma-rays and neutrons, respectively.     

An Adaptive Control Concept of Reactor Power Distribution

A new method has been developed for obtaining the β-ray self-absorption correction factor of a Au-foil with uniformly distributed A-ray source. The method is essentially to average the β-ray self-absorption correction factor of Au-foil packs composed of non-activated Au-layers and an activated Au-layer placed at the various depths.

The β-ray self-absorption correction factors of the Au-foil with thicknesses ranging 0.15–80mg/cm² were measured within an experimental error of less than 0.5%. Between the thicknesses of 0 and 10mg/cm², a sharp decrease was observed. In the range of thickness below 45mg/cm² the present experimental results diverge from the currently used formula (f _a = (l–e ^μt)/μt), but are in good agreement with a semi-empirical expression (f _a = (l/2–E ₃(^μt)/μt), formulated on the assumption of isotropic penetration of β-rays.

The escape probability of β-rays emitted from a certain depth of Au-foil was also measured as a function of source depth. By using this escape probability it is possible to obtain the β-ray self-absorption correction factors for any concave source distribution through the thickness of a foil, a situation commonly encountered especially in the foil irradiated with epi-thermal neutrons.     

Calculation of the cross-sections for fast neutrons and gamma-rays in concrete shields

This work is concerned with the theoretical calculation of both the total mass attenuation coefficients (μ/ρ) for gamma rays and the effective removal cross-sections (Σ_R) for fast neutrons. Calculations were carried out for four types of concrete of different densities namely (1) dolomite–sand ρ=2.5 g.cm⁻³, (2) barite–barite ρ=3.49 g.cm⁻³, (3) magnetite–limonite ρ=3.6 g.cm⁻³ and (4) ilmenite–ilmenite ρ=3.69 g cm⁻³. The total mass attenuation coefficients have been calculated at energies from 10 keV to 1 GeV using the XCOM (version 3.1) computer program and cross- section database for elements from Z=1 to 100. Also the effective removal cross-section have been calculated using the elemental composition of the concrete mixes. The calculated values are widely needed and used as a database for radiation shielding design of research reactors, power station and particle accelerators. The results were displayed in tables and figures. Calculated results were compared with those previously measured and a reasonable agreement was found between them.     

Gamma-ray attenuation method as an efficient tool to investigate soil bulk density spatial variability

The spatial variability of soil bulk density (ρ_b) was measured by using the volumetric ring method (VRM) and the gamma-ray attenuation method (GAM). Collimated radiation from 3.7 GBq of ²⁴¹Am was used to evaluate the soil mass attenuation coefficient and its bulk density. Circular lead collimators were adjusted and aligned between source (D = 1, 2 and 3 mm) and detector (D = 4.5 mm). Results of GAM for average ρ_b provided good agreement with the corresponding values obtained gravimetrically. Variations in bulk density for different collimator dimensions can be attributed to multiple scattering after photons interaction with soil, mainly for 3 mm collimator size. The best result of ρ_b by the nuclear technique was obtained when ρ_b represents an average of the measurements for collimators of 1 and 2 mm. Another cause for the differences in ρ_b by GAM and VRM is the heterogeneity of soil when the collimated beam can interact with stones or large air-filled holes or channels present in the sample. Therefore, the pattern of spatial variability obtained by VRM was confirmed by GAM for all collimator sizes. This result is a good indication that GAM can be used with success to analyze soil spatial variability.     

克量粉末样品中低比活度放射性核素的γ能谱分析

报道了缺少小直径圆盘状体标准源条件下,快速简便地测定固体粉末样品中多种人工放射性核素比活度的γ能谱分析方法。取1g左右样品,装在特制的塑料盒中,形成Φ26mm×3mm的小圆盘状。由于样品的比活度较低,采用低源距(1cm,记为H1)测量。由于没有与样品圆盒相同形式的体标准源,利用直径26mm的60Co平面标准源在高源距(25cm,记为H25)测定若干放射性核素平面源的活度及峰效率,再测定H1的平面源峰效率。然后源距在H1基础上逐次增加1mm以测量平面源的峰效率,拟合峰效率随距离变化的函数,模拟得到厚度为3mm圆盘体标准源的峰效率。最终测定了样品比活度(包含自吸收校正)。与样品中两种核素的放化分析结果比对,测定结果偏差≤4%。     

Measurement of ~(232)Th(n,2n)~(231)Th reaction cross-sections at neutron energies of 14.1 MeV and 14.8 MeV using neutron activation method

In this study, the activation cross-sections were measured for ~(232)Th(n,2n)~(231)Th reactions at neutron energies of 14.1 and 14.8 MeV, which were produced by a neutron generator through a T(d,n)~4He reaction. Induced gamma-ray activities were measured using a low background gamma ray spectrometer equipped with a high resolution HPGe detector. In the cross-section calculations, corrections were made regarding the effects of gamma-ray attenuation, dead-time, fluctuation of the neutron flux, and low energy neutrons. The measured cross-sections were compared with the literature data, evaluation data(ENDF-B/VII.1, JENDL-4.0 and CENDL-3.1), and the results of the model calculation(TALYS1.6).     

Determination of γ-spectrometric parameters for some chloride salt solutions

The mass attenuation coefficients, atomic cross-sections and effective atomic numbers are determined for LiCl, NaCl and KCl solutions with contents up to 0.2 g cm⁻³.The energies used are 0.511, 0.662 and 1.274 keV. The combined results were employed to determine the mass attenuation coefficients and atomic cross-sections of the salts and water in the solutions. The results are compared with the calculated values from elemental data. It was found that the mass attenuation coefficients and corresponding total cross-sections are, in general, less from measurement than from tabulated data, either for water or for the chloride salts.     

Determination of the plutonium mass in containers with plutonium dioxide according to the γ radiation from spontaneous-fission products

A nondestructive method of determining the mass of large plutonium samples which is based on measuring their characteristic γ radiation is presented. The mass and the isotopic composition of the plutonium were determined according to a single measured γ spectrum in two energy ranges: the middle range, where the isotopic composition of plutonium is determined from the lines of the plutonium isotopes, and high-energy, where the radiation of the products of the spontaneous fission of plutonium gives information about the mass of the sample. The dependence of the counting rate in the peaks of the fission products on the effective mass of ²⁴⁰Pu was calibrated according to measurements performed with standard samples of the enterprise. As a result of the measurements, corrections were made for the self-absorption of γ radiation and induced fission, which were calculated by the Monte Carlo method. The error in determining the plutonium dioxide mass is 3–10% for containers with different cool-down times and isotopic composition of plutonium.     

The dose rate effect and the homogeneity of radio-oxidation of plastics

The homogeneity of the radio-oxidation of plastics in different depths from the surface has been determined by measuring the thermo-oxidative stability (oxidative induction time – OIT) of irradiated samples. Two materials have been studied: a fire retarding EPR/EVA cable sheathing compound with the thickness of 4.4 mm and a high-density polyethylene (HDPE) (20 mm thick) which was studied for potential use as a material for radioactive waste disposal containers. Both materials have been irradiated using ⁶⁰Co gamma-ray source at different dose rates in the interval from 8.5 to 8550 Gy/h. Irradiated samples have been cut into very thin slices and the thermo-oxidative stability (OIT) has been measured using differential scanning calorimeter. In this way the dependence of OIT values on the distance from the surface has been obtained for both samples and at applied dose rates.     

Simulating gamma-ray energy resolution in scintillators due to electron-hole pair statistics

The best-possible limit to gamma-ray energy resolution in scintillators is given by the statistics of the number of electron-hole pairs produced by an incident gamma-ray, characterized by the Fano factor. The Fano factor is primarily controlled by the inelastic scattering during the electron cascade, which could be modeled by Monte Carlo simulation. Commonly used radiation transport codes do not follow the electrons to low enough energies to calculate electron-hole pair distributions. A Monte Carlo simulation for inelastic electron scattering is introduced based on cross-sections derived from data measured by Electron Energy-Loss Spectroscopy (EELS) for fast electrons. This inelastic scattering model was incorporated into the radiation transport code Penelope so that it could accurately count the number of electron-hole pairs produced by a gamma-ray. The Fano factor was calculated for the scintillators cerium fluoride (CeF₃) and lutetium oxyorthosilicate (Lu₂SiO₅).