Hyper pure quartz as a promising material for retrospective and radiation processing dosimetry using ESR technique.

Samples of highly purified naturally occurring silicon dioxide (SiO(2)) commercially named clear fused quartz (CFQ), both in bulk and in powder form were exposed to 4 and 8 MeV electrons from a linear accelerator (radiotherapy unit) at doses between 0.5 and 60 Gy (minimum and maximum attainable dose, respectively). After irradiation, the samples were subjected to electron spin resonance (ESR) analysis readout to test their use as a mean of measuring absorbed doses in the samples. A good linearity between the absorbed dose (from 2 Gy up to the maximum dose applied of 60 Gy) and the ESR intensity of the samples was observed while a poor linearity at low doses (between the absorbed dose from 0.5 to 10 Gy and the ESR intensity) is seen. The background intensity in the case of the powder was higher than for the cylindrical samples and the ESR intensity was also lower in the former case, which gave rise to a dose curve response with a lower slope for the powders. No energy dependence was apparent for electrons of 4 and 8 MeV energy. The thermal fading of CFQ at room temperature showed a relatively sharp decay of approximately 30% over a period of 10 days after which the decay rate was negligible for a measured period of 80 days. The results in this study plus the previous work done (gamma irradiation of the samples) by the authors suggest that the materials could, by using the ESR technique, be a suitable candidate for alternative dose measurements in radiation processing and retrospective dosimetry.     

Thermoluminescence characteristics of Nd-doped SiO2 optical fibers irradiated with the 60Co gamma rays

Thermoluminescence (TL) properties (radiation sensitivity, dose response, signal fading) of Nd-doped SiO₂ optical fibers irradiated with 1.25 MeV photons to 1–50 Gy were studied. The peak of the glow curve is around 190 °C regardless of the dose. The dose response is linear up to 50 Gy. The radiation sensitivity is 219 nC mg⁻¹ Gy⁻¹. The fiber can be a potential candidate for photon radiotherapy dosimetry due to its high radiation sensitivity, linear dose response in a wide range, slow fading, and high spatial resolution due to the small size of the fiber.     

Concrete enclosure for shielding a neutron source

In the aim to design a shielding for a 0.185 TBq ²³⁹PuBe isotopic neutron source several Monte Carlo calculations were carried out using MCNP5 code. First, a point-like source was modeled in vacuum and the neutron spectrum and ambient dose equivalent were calculated at several distances ranging from 5 cm up to 150 cm, these calculations were repeated modeling a real source, including air, and a 1×1×1 m³ enclosure with 5, 15, 20, 25, 30, 50 and 80 cm-thick Portland type concrete walls. At all the points located inside the enclosure neutron spectra from 10⁻⁸ up to 0.5 MeV were the same regardless the distance from the source showing the room-return effect in the enclosure, for energies larger than 0.5 MeV neutron spectra are diminished as the distance increases. Outside the enclosure it was noticed that neutron spectra becomes “softer” as the concrete thickness increases due to reduction of mean neutron energy. With the ambient dose values the attenuation curve in terms of concrete thickness was calculated.     

Natural gamma-ray spectrometry as a tool for radiation dose and radon hazard modelling

We reviewed the calibration procedures of gamma-ray spectrometry with particular emphasis to factors that affect accuracy, detection limits and background radiation in field measurements for dosimetric and radon potential mapping. Gamma-ray spectra were acquired in western Liguria (Italy). The energy windows investigated are centred on the photopeaks of ²¹⁴Bi (1.76 MeV), ²⁰⁸Tl (2.62 MeV) and ⁴⁰K (1.46 MeV). The inferred absorbed dose rate and the radon flux are estimated to be lower than 60 nGy h⁻¹ and 22 Bq m⁻² h⁻¹, respectively.     

LiSr4(BO3)3:Ce3+ phosphor as a new material for ESR dosimetry

LiSr₄(BO₃)₃:0.01Ce³⁺ phosphor was investigated to assess its potential as a material for measurements of radiotherapeutic doses with electron spin resonance (ESR). The ESR spectrum of the phosphor irradiated with ⁶⁰Co features five ESR signals. An isochronal annealing experiment has shown that the strongest of these signals is associated with the same trap center as the 473 K peak on the TL glow curve of this material. The dose–response is linear in the studied range from 0.89 to 90.30 Gy. Fading of the signal was also investigated.     

The thermoluminescence response of doped SiO2 optical fibres subjected to photon and electron irradiations

Modern linear accelerators, the predominant teletherapy machine in major radiotherapy centres worldwide, provide multiple electron and photon beam energies. To obtain reasonable treatment times, intense electron beam currents are achievable. In association with this capability, there is considerable demand to validate patient dose using systems of dosimetry offering characteristics that include good spatial resolution, high precision and accuracy. Present interest is in the thermoluminescence response and dosimetric utility of commercially available doped optical fibres. The important parameter for obtaining the highest TL yield during this study is to know the dopant concentration of the SiO₂ fibre because during the production of the optical fibres, the dopants tend to diffuse. To achieve this aim, proton-induced X-ray emission (PIXE), which has no depth resolution but can unambiguously identify elements and analyse for trace elements with detection limits approaching μg/g, was used. For Al-doped fibres, the dopant concentration in the range 0.98–2.93 mol% have been estimated, with equivalent range for Ge-doped fibres being 0.53–0.71 mol%. In making central-axis irradiation measurements a solid water™ phantom was used. For 6-MV photons and electron energies in the range 6, 9 and 12 MeV, a source to surface distance of 100 cm was used, with a dose rate of 400 cGy/min for photons and electrons. The TL measurements show a linear dose–response over the delivered range of absorbed dose from 1 to 4 Gy. Fading was found to be minimal, less than 10% over five days subsequent to irradiation. The minimum detectable dose for 6-MV photons was found to be 4, 30 and 900 μGy for TLD-100 chips, Ge- and Al-doped fibres, respectively. For 6-, 9- and 12-MeV electron energies, the minimum detectable dose were in the range 3–5, 30–50 and 800–1400 μGy for TLD-100 chip, Ge-doped and Al-doped fibres, respectively.     

Radiochemical determination of cross sections of α-particle induced reactions on 192Os for the production of the therapeutic radionuclide 193mPt

For determination of cross sections of α-particle induced reactions on 99.65% enriched ¹⁹²Os, the methods for electrolytic preparation of thin samples and radiochemical separation of radioplatinum were optimized. The excitation functions of the ¹⁹²Os(α,n)^195mPt and ¹⁹²Os(α,3n) ^193mPt reactions were measured from 20 to 39 MeV. The cross section of the latter reaction reaches a maximum value of about 1.5 b at an energy around 36 MeV. The results of nuclear model calculations using the codes TALYS and STAPRE agreed well with the measured data. The optimum energy range for the production of no-carrier-added ^193mPt (T_1/2=4.33 d) was found to be E_α=40→30 MeV. The thick target yield amounts to 10 MBq/μA h and a possible batch yield of 2 GBq should be sufficient for Auger electron therapy on a wide scale.     

Reproducibility of fascicle length and pennation angle of gastrocnemius medialis in human gait in vivo

The purpose of the current study was to examine the reproducibility of fascicle length and pennation angle of gastrocnemius medialis while human walking. To the best of our knowledge, this is the first study of the reproducibility of fascicle length and pennation angle of gastrocnemius medialis in vivo during human gait. Twelve males performed 10 gait trials on a treadmill, in 2 separate days. B-mode ultrasonography, with the ultrasound probe firmly adjusted in the transverse and frontal planes using a special cast, was used to measure the fascicle length and the pennation angle of the gastrocnemius medialis (GM). A Vicon 624 system with three cameras operating at 120 Hz was also used to record the ankle and knee joint angles. The results showed that measurements of fascicle length and pennation angle showed high reproducibility during the gait cycle, both within the same day and between different days. Moreover, the root mean square differences between the repeated waveforms of both variables were very small, compared with their ranges (fascicle length: RMS = ～3 mm, range: 38–63 mm; pennation angle: RMS = ～1.5°, range: 22–32°). However, their reproducibility was lower compared to the joint angles. It was found that representative data have to be derived by a wide number of gait trials (fascicle length ～six trials, pennation angle more than 10 trials), to assure the reliability of the fascicle length and pennation angle in human gait.     

Application of jade samples for high-dose dosimetry using the EPR technique

The dosimeter characteristics of jade samples were studied for application in high-dose dosimetry. Jade is the common denomination of two silicates: jadeite and actinolite. The EPR spectra of different jade samples were obtained after irradiation with absorbed doses of 100 Gy up to 20 kGy. The jade samples present signals that increase with the absorbed dose (g-factors around 2.00); they can be attributed to electron centers. The EPR spectra obtained for the USA jade samples and their main dosimetric properties as reproducibility, calibration curves and energy dependence were investigated.     

Using medical accelerators and photon activation to determine Sr/Ca concentration ratios in teeth

This paper describes a photon activation method, studied by using two medical accelerators (energies: 15 and 18 MeV) as photon sources, for determining Sr and Ca levels and Sr/Ca ratios in tooth samples. The radionuclides formed by various photonuclear reactions were measured and identified using a gamma-spectrometry with HPGe detection system. The yields of the corresponding photonuclear reactions and the detection sensitivities for the alkaline earth metals (e.g., Ca, Sr) were surveyed and estimated in relation to the radiation dose. The minimum detectable amount of Sr was estimated to be less than 1 μg g⁻¹, allowing the Sr/Ca ratios in teeth to be determined conveniently. The Sr/Ca ratios in deciduous and permanent tooth samples obtained from local dental clinics were 0.390 and 0.565 mg g⁻¹, respectively. This photon activation method of determining Sr/Ca ratio in bones and teeth using medical accelerators for cancer treatment is thought to be useful also in biological and archaeological studies.     

Alpha particle spectrometry with Lexan polycarbonate using combined layer removal by ethylenediamine solution and electrochemical etching

By applying a pre-etching (PE) process to remove a given surface layer followed by electrochemical etching with potassium hydroxide/ethanol/water (PEW) at 25°C it is possible to differentiate between different energies (1 MeV, 2 MeV, 3 MeV, 4 MeV, and 5 MeV) of alpha particles. The method of layer removal is based on the chemical reaction between polycarbonate and ethylenediamine. The relationship between the thickness of the removed layer and the concentration of ethylenediamine solution for different times of soaking was determined. For each concentration of ethylenediamine in water, the bulk etching rate, V_B, was measured by the gravimetric method.     

Distribution of radionuclides in soil samples from a petrified wood forest in El-Qattamia,Cairo, Egypt

The concentrations and distribution of radionuclides in a petrified wood forest in El-Qattamia have been determined using high-resolution gamma spectrometry to evaluate the environmental radioactivity. The mean activity concentrations of ²³⁸U, ²³²Th and ⁴⁰K were 65.26±12.99, 23.66±0.95 and 146.33±1.50 Bq kg⁻¹, respectively.Data of the soil samples show evidence of possible deposition and accumulation of ¹³⁷Cs. The mean activity concentration of ¹³⁷Cs in the soil samples was 4.37±0.16 Bq kg⁻¹ with a range of 0.00–35.70 Bq kg⁻¹. The measured activity concentration range of ¹³⁷Cs was compared with reported ranges in the literature from some of the other locations in the world. The radium-equivalent, dose rate in air and annual effective dose rate were evaluated. The mean activity concentrations of the γ-ray emissions from radionuclides in El-Qattamia petrified wood forest region were relatively low.     

Near-threshold 7Li(p,n)7Be neutrons on the practical conditions using thick Li-target and Gaussian proton energies for BNCT

The near threshold ⁷Li(p,n)⁷Be neutrons generated by incident proton energy having Gaussian distribution with mean energies from 1.85 to 1.95 MeV, were studied as a practical neutron source for BNCT wherein an RFQ accelerator and a thick Li-target are used. Gaussian energy distributions with the standard deviation of 0, 10, 20 and 40 keV for mean proton energies from 1.85 to 1.95 MeV were surveyed in 0.01 MeV increments. A thick liquid Li-target whose dimensions were established in our previous experiments (i.e., 1 mm-thick with 50 mm width and 50 mm length) was considered in this study. The suitable incident proton energy and physical dimensions of Pb layer which serves as a gamma absorber and a Polyethylene layer which is used as a BDE were surveyed by means of the concepts of TPD. Dose distribution were calculated by using MCNP5. A proton beam with mean energy of 1.92 MeV and a Gaussian energy distribution with a standard deviation of 20 keV at a current of 10 mA was selected from the viewpoint of irradiation time and practically achievable proton current. The suitable thicknesses of Pb gamma absorber was estimated to be about 3 cm. The estimated thickness of the polyethylene BDE was about 24 mm for an ideal proton current of 13 mA, and was 18 mm for a practical proton current of 10 mA.     

Dose profile modeling of Idaho National Laboratory's active neutron interrogation laboratory

A new laboratory has been commissioned at Idaho National Laboratory for performing active neutron interrogation research and development. The facility is designed to provide radiation shielding for deuterium–tritium (DT) fusion (14.1 MeV) neutron generators (2×10⁸ n/s), deuterium–deuterium (DD) fusion (2.5 MeV) neutron generators (1×10⁷ n/s), and ²⁵²Cf spontaneous fission neutron sources (6.96×10⁷ n/s, 30 μg). Shielding at the laboratory is comprised of modular concrete shield blocks 0.76 m thick with tongue-in-groove features to prevent radiation streaming, arranged into one small and one large test vault. The larger vault is designed to allow operation of the DT generator and has walls 3.8 m tall, an entrance maze, and a fully integrated electrical interlock system; the smaller test vault is designed for ²⁵²Cf and DD neutron sources and has walls 1.9 m tall and a simple entrance maze. Both analytical calculations and numerical simulations were used in the design process for the building to assess the performance of the shielding walls and to ensure external dose rates are within required facility limits. Dose rate contour plots have been generated for the facility to visualize the effectiveness of the shield walls and entrance mazes and to illustrate the spatial profile of the radiation dose field above the facility and the effects of skyshine around the vaults.     

A 238Pu irradiator for exposure of cultured cells with alpha-radiation: Construction,calibration and dosimetry

An alpha-particle irradiator that can facilitate investigations of alpha-radiation effects on human cells in radiation protection, carcinogenesis and radioimmunotherapy was constructed. The irradiator was based on a 1.3 GBq ²³⁸Pu source, housed in a stainless steel tube flushed with helium. Radiation provided by ²³⁸Pu consists mainly of alpha-particles with energy of 5.5 MeV. The alpha-particle fluence and energy spectra were measured with a silicon semiconductor detector. Monte Carlo simulations were used to estimate the mean number of alpha-particles and the mean absorbed alpha-particle dose to cells for various irradiation times and distances between cells and source. There was a linear dependence between exposure time and alpha-particle fluence for exposure times above 1 s. The alpha-particle activity concentration varied with a factor 2.7 over the source area, while the variation in energy peak position was <4%. At the cell nucleus position and with a distance of 45 mm between the source and the mylar dish surface, the α-fluence was 4.6×10⁴ counts/(mm² s), the average incident α-particle energy was 2.5 MeV and the average linear energy transfer was 167 keV/μm. The average dose rate to the cells, with 5 μm diameter nucleus, was 1.2 Gy/s. The ²³⁸Pu α-particle irradiator is feasible for irradiation of cells and it can be used for studies of both direct effects and bystander effects of alpha-radiation.     

Emphysema quantification and lung volumetry in chest X-ray equivalent ultralow dose CT – Intra-individual comparison with standard dose CT

ObjectivesTo determine whether ultralow dose chest CT with tin filtration can be used for emphysema quantification and lung volumetry and to assess differences in emphysema measurements and lung volume between standard dose and ultralow dose CT scans using advanced modeled iterative reconstruction (ADMIRE).Methods84 consecutive patients from a prospective, IRB-approved single-center study were included and underwent clinically indicated standard dose chest CT (1.7 ± 0.6 mSv) and additional single-energy ultralow dose CT (0.14 ± 0.01 mSv) at 100 kV and fixed tube current at 70 mAs with tin filtration in the same session. Forty of the 84 patients (48%) had no emphysema, 44 (52%) had emphysema. One radiologist performed fully automated software-based pulmonary emphysema quantification and lung volumetry of standard and ultralow dose CT with different levels of ADMIRE. Friedman test and Wilcoxon rank sum test were used for multiple comparison of emphysema and lung volume. Lung volumes were compared using the concordance correlation coefficient.ResultsThe median low-attenuation areas (LAA) using filtered back projection (FBP) in standard dose was 4.4% and decreased to 2.6%, 2.1% and 1.8% using ADMIRE 3, 4, and 5, respectively. The median values of LAA in ultralow dose CT were 5.7%, 4.1% and 2.4% for ADMIRE 3, 4, and 5, respectively. There was no statistically significant difference between LAA in standard dose CT using FBP and ultralow dose using ADMIRE 4 (p = 0.358) as well as in standard dose CT using ADMIRE 3 and ultralow dose using ADMIRE 5 (p = 0.966). In comparison with standard dose FBP the concordance correlation coefficients of lung volumetry were 1.000, 0.999, and 0.999 for ADMIRE 3, 4, and 5 in standard dose, and 0.972 for ADMIRE 3, 4 and 5 in ultralow dose CT.ConclusionsUltralow dose CT at chest X-ray equivalent dose levels allows for lung volumetry as well as detection and quantification of emphysema. However, longitudinal emphysema analyses should be performed with the same scan protocol and reconstruction algorithms for reproducibility.     

The natBr(p,x)73,75Se nuclear processes: a convenient route for the production of radioselenium tracers relevant to amino acid labelling

A possible route for the production of no-carrier-added (n.c.a.) ⁷³Se (T_1/2=7.1 h) and ⁷⁵Se (120 d) is introduced. d,l-2-Amino-4-([⁷³Se]methyl-seleno) butanoic acid (d,l-[⁷³Se]selenomethionine) with an overall radiochemical yield of >40% could be prepared via a 3-step polymer-supported synthesis after successful separation of ⁷³Se from KBr targets. Excitation functions for the ^natBr(p,x) ^72,73,75Se processes were measured from threshold up to 100 MeV utilizing pellets of pressed KBr. Targets were irradiated at the NAC cyclotron with proton beams having primary energies of 40.4, 66.8 and 100.9 MeV. The calculated ⁷³Se yield (EOB) for 1 h irradiation in 1 μA of beam at the optimum proton energy range of 62→42 MeV is 81.4 MBq (2.2 mCi), and the calculated ⁷⁵Se yield (EOB) for the overall range 62 MeV→threshold for the same irradiation conditions is 0.97 MBq (0.026 mCi).     

A comparison of HDR near source dosimetry using a treatment planning system,Monte Carlo simulation,and radiochromic film

This study aimed to investigate the high-dose rate Iridium-192 brachytherapy, including near source dosimetry, of a catheter-based applicator from 0.5 mm to 1 cm along the transverse axis. Radiochromic film and Monte Carlo (MC) simulation were used to generate absolute dose for the catheter-based applicator. Results from radiochromic film and MC simulation were compared directly to the treatment planning system (TPS) based on the American Association of Physicists in Medicine Updated Task Group 43 (TG-43U1) dose calculation formalism. The difference between dose measured using radiochromic film along the transverse plane at 0.5 mm from the surface and the predicted dose by the TPS was 24%±13%. The dose difference between the MC simulation along the transverse plane at 0.5 mm from the surface and the predicted dose by the TPS was 22.1%±3%. For distances from 1.5 mm to 1 cm from the surface, radiochromic film and MC simulation agreed with TPS within an uncertainty of 3%. The TPS under-predicts the dose at the surface of the applicator, i.e., 0.5 mm from the catheter surface, as compared to the measured and MC simulation predicted dose. MC simulation results demonstrated that 15% of this error is due to neglecting the beta particles and discrete electrons emanating from the sources and not considered by the TPS, and 7% of the difference was due to the photon alone, potentially due to the differences in MC dose modeling, photon spectrum, scoring techniques, and effect of the presence of the catheter and the air gap. Beyond 1 mm from the surface, the TPS dose algorithm agrees with the experimental and MC data within 3%.