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PURPOSE With improvements in positron emission tomography/magnetic resonance imaging (PET/MRI) over the last decade, there is a need to investigate the projected annihilation (shine-through) artifact and resolution impact for different PET radiopharmaceuticals, magnetic field (MF) strengths, and tissues. METHODS The GATE Monte Carlo (MC) simulation was used to simulate the annihilation distribution of positrons in different tissues and MFs. The positron distribution was studied in magnetic field (MF) intensities up to 15 T for 11 C, 13 N, 15 O, 18 F, 68 Ga, and 82 Rb. Moreover, the image quality in terms of the occurrence of projected annihilation artifacts was investigated using the 4D anthropomorphic digital extended cardiac-torso (XCAT) phantom. RESULTS Positron ranges were restricted across the directions perpendicular to the MF, but no change along the direction of the MF was detected. The projected annihilation artifacts were observed with the presence of MF in the sagittal and coronal view of PET images prepared from the XCAT phantom. The intensity of artifact was constant in MFs higher than 3 T. The significant effect of the MF on resolution improvement was observed in soft tissue for 68 Ga in 7 T and 82 Rb in 3 and 7 T, while higher MFs have no impact on resolution. The improvement of resolution in the lung tissue was observed for the medium- and high-energy radionuclides in 7 T MF. CONCLUSION The MF can create the projected annihilation artifact in the boundary of air cavities and other tissues for medium- and high-energy radionuclides especially for 68 Ga in clinical studies. In addition, the strength of the MFs more than 3 T was ineffective on the intensity of the projected annihilation artifact. In a clinical PET/MR scanner, MF has remarkable spatial resolution improvement in lung tissue, especially for medium- and high-energy radionuclides, and negligible effect in bone and soft tissue for most radionuclides.

Purpose: There are few areas in the world where ionizing radiation (IR) dose received by the public from radon gas and other radioactive elements are much higher than recommended limits. Telomere length is a potential biomarker of genomic instability due to oxidative stress from IR exposure. In this study, we investigated the impact of chronic exposure to environmental IR on relative telomere length (RTL) in white blood cells (WBC) among inhabitants of high background radiation areas (HBRA) of Ramsar, Iran. Materials and methods: One hundred and four individuals from HBRAs of Ramsar and 104 age-matched subjects from normal background radiation areas (NBRA) of Iran were enrolled in the study. The RTLs of WBC DNA samples were measured by a quantitative polymerase chain reaction (q-PCR) approach. Results: Mean RTL in HBRA and NBRA groups did not show any significant difference (1.21 ± 0.71 versus 1.22 ± 0.66, p = .306). After controlling for all demographic variables, less than 1% of the variances in RTL values were related to background radiation exposure. Conclusion: We conclude that chronic exposure to natural IR has no statistically significant effect on RTL among the inhabitants of HBRAs of Ramsar compared with a control group.

Purpose Our current study aims to consider the image biomarkers extracted from the MRI images for exploring their effects on glioblastoma multiforme (GBM) patients' survival. Determining its biomarker helps better manage the disease and evaluate treatments. It has been proven that imaging features could be used as a biomarker. The purpose of this study is to investigate the features in MRI and clinical features as the biomarker association of survival of GBM. Methods 55 patients were considered with five clinical features, 10 qualities pre-operative MRI image features, and six quantitative features obtained using BraTumIA software. It was run ANN, C5, Bayesian, and Cox models in two phases for determining important variables. In the first phase, we selected the quality features that occur at least in three models and quantitative in two models. In the second phase, models were run with the extracted features, and then the probability value of variables in each model was calculated. Results The mean of accuracy, sensitivity, specificity, and area under curve (AUC) after running four machine learning techniques were 80.47, 82.54, 79.78, and 0.85, respectively. In the second step, the mean of accuracy, sensitivity, specificity, and AUC were 79.55, 78.71, 79.83, and 0.87, respectively. Conclusion We found the largest size of the width, the largest size of length, radiotherapy, the volume of enhancement, volume of nCET, satellites, enhancing margin, and age feature are important features.

Purpose: One of the promising radiosensitizers is the ultrasmall gold nanoparticle (GNP) with a hydrodynamic diameter vβ3) in the murine ALTS1C1 glioma cell line. Materials and methods: We investigated the uptake, toxicity and radiosensitivity of GNP–PEG–cRGDfKs in ALTS1C1 cells exposed to protons, kilovoltage photons and megavoltage photons. The in vitro uptake and toxicity of GNPs in the hepatocytes and Kupffer cells were assessed for murine AML12 hepatocyte and RAW 264.7 macrophage cell lines. The in vivo biodistribution of GNPs in the ALTS1C1 tumor model was tested using the inductively coupled plasma mass spectrometry. Results: Results indicated GNPs accumulated in the cytoplasm with negligible toxicity for a moderate concentration of GNPs. Observed sensitizer enhancement ratios and dose enhancement factors are 1.21–1.66 and 1.14–1.33, respectively, for all radiations. Conclusion: Ultrasmall GNP–PEG–cRGD can be considered as a radiosensitizer. For radiotherapy applications, the delivery method should be developed to increase the GNP uptake in the tumor and decrease the uptakes in undesirable organs.

Nuclear reactions of high-energy protons with treatment equipment, air, and patient tissue during proton therapy generate residual radioactivity and secondary particles including protons, deuterons, alphas, and neutrons. The most up-to-date versions of INCL++ (v5.2.9), TALYS (v1.8), EMPIRE (v3.2.2 Malta), and ALICE/ASH were used in this study to calculate the excitation functions of proton-induced reactions with carbon, nitrogen, oxygen, aluminum, calcium, iron, nickel, copper, zinc, tin, tungsten, and lead nuclei. The cross sections of different nuclear reaction mechanisms, gamma particles, and residual radionuclides were calculated. The obtained results were compared with available experimental data and the ENDF/B-VII.1.

The purpose of this study is to perform absorbed dose calculations based on Monte Carlo simulations for a novel beta emitter bioglass Ho-166 seed which is proposed for treating small hepatocellular carcinomas (HCCs). The bioactive glass seed has been developed by use of the sol–gel method. Monte Carlo simulations were carried out for the seed using the version 5 of the (MCNP) Monte Carlo radiation transport code to investigate the dosimetric parameters recommended by the AAPM Task Group 60 (TG-60). Dose distributions due to the beta and photon radiation were obtained at different radial distances surrounding the source. The dose rate in water at the reference point was calculated to be 6.71±0.4 cGy h−1 μCi−1. The anisotropy function values ranging from 0.745 to 1.928 were obtained for radial distances of 0.3–8 mm and polar angles of 0°–90°. The 166Ho seed source can deliver high radiation doses to the tumor, while the short range of the beta particles limits damage to the adjacent normal tissue.

We here presented a method for zirconium-89 production using a low-energy cyclotron. The half-life of 89Zr (T1/2 = 78.4 h) is suitable for immunoPET. 89Zr was produced via the 89Y(p, n)89Zr nuclear process at 20 μA current and 15 → 5 MeV protons beam (20 min). Yield of about 60.77 MBq 89Zr per μAh was experimentally obtained. Also, excitation functions and theoretical yields via 89Y(p, n)89Zr, 89Y(d, 2n)89Zr, natSr(α, xn)89Zr and 90Zr(n, 2n)89Zr reactions were calculated by EMPIRE (version 3.1 Rivoli), ALICE/ASH (pre-equilibrium and equilibrium) and TALYS-1.26 (direct, pre-equilibrium, and equilibrium) codes and TENDL-2010 database. The obtained results have been discussed and compared with the available experimental data.

Background: Today, cancer treatment is one of challenges in health and treatment system of any country. With quick development of radiation therapy and chemotherapy in recent years, this question is brought up that which method is more perfect to treat cancer. Methods: Radiation therapy is consisted of several methods including external radiation therapy (e.g. CyberKnife, IMRT, IGRT, SBRT, BNCT, Vero, IORT, Gamma knife, Tomotherapy and Volumetric modulated arc therapy (VMAT)) and internal radiation therapy (e.g. brachytherapy, endoradiotherapy (target radionuclide therapy): radioimmunotherapy and peptide receptor radionuclide therapy) that any one has a special function in cancerous cells destroy. Results: Chemotherapy has still its place in cancer treatment and is used as a current and worldwide method. Nowadays, radiation therapy evolves with a surprising growth. By abandoning Cobalt-60, linear and circular accelerators which are able to accelerate particles, could be used for cancer treatment. CyberKnife, IMRT, IGRT, SBRT, VMAT, tomotherapy and Vero are the most advanced methods used to radiation therapy in all angles which had minimum mistake in radiation to the near intact tissues. Today, brachytherapy by increasing production of treatment radioisotopes by cyclotron and reactor is used in the world. Endoradiotherapy (Target radionuclide therapy): Radioimmunotherapy and peptidereceptor radionuclide therapy are two considerable methods in cancer treatment. Conclusion: In this review, we consider principals and application of radiation therapy in the early 21st century. © 2011 Bentham Science Publishers.

A method was suggested for the cyclotron production of 88Y with liquid-liquid extraction. The sedimented natSrCO3 target was irradiated with 18-MeV protons at current of 20 μA for 10 h. The 88Y yields of about 1.326 MBq μA−1 h−1 were experimentally obtained. Solvent extraction of no-carrier-added 88Y from irradiated strontium carbonate target in the hydrochloric acid solution was studied using di(2-ethylhexyl) hydrogen phosphate (HDEHP). The optimum separation was achieved in the system n-hexane/10% HDEHP-0.1 M HCl. Yttrium radionuclides were recovered from the HDEHP phase by stripping with 60 ml of 9 M HC1. Also, excitation functions of the proton, deuteron, and α-particle induced reactions of 89Y, 88Sr, natSr, and natRb were determined using computer codes and compared to the existing data.

In this study, design and fabrication of 103Pd brachytherapy seed was investigated. The excitation functions of 103Rh(p,n) 103Pd and 103Rh(d,2n)103Pd reactions were calculated using EMPIRE (version 3.1 Rivoli), ALICE/ASH and TALYS-1.2 codes, the TENDL-2010 database and compared with the published data. Production of 103Pd was done via 103Rh(p,n)103Pd nuclear reaction. The target was bombarded with 18 MeV protons at 200 μA beam current for 15 h. After irradiation and radiochemical separation of the electroplated rhodium target, the optimum condition for absorption of 103Pd into Amberlite®IR-93 resin was achieved at 0.5 M HCl. Version 5 of the (MCNP) Monte Carlo radiation transport code was employed to calculate the dosimetric parameters around the 103Pd brachytherapy seed. Finally the calculated results were compared with published results for other commercial sources. © 2011 Elsevier Ltd. All rights reserved.

The excitation functions for the 72Ge(p,n)72As and natGe(p,xn)72 As reactions were calculated by ALICE/91, ALICE/ASH, and TALYS-1.2 codes and compared to the existing data. Deposition of GeO2 on a Cu substrate was carried out using two special sedimentation methods for the production of 72As, and a simple procedure for As separation was described. The targets were checked by SEM and thermal shock test. The optimal thicknesses of the target was substantiated, and theoretical calculation of the production yield was made for the 72Ge(p,xn)72As reaction.

The nuclear-physical properties of 45Ti (half-life 3.08 h, 85% β+-decay, 15% electron capture, daughter isotope 45Sc) make it suitable for positron emission tomography. 45Ti can be produced at a small cyclotron in a high yield, 450 MBq μA−1 h−1, using the 45Sc(p,n)45Ti nuclear reaction on a natural scandium target. The excitation functions for the 45Sc(p,n)45Ti reaction were calculated by ALICE/91, ALICE/ASH, and TALYS-1.2 codes and compared to the existing data. Deposition of Sc2O3 on a Cu substrate was carried out by two sedimentation methods for the production of 45Ti, and a simple method for radiochemical separation of radiotitanium was suggested. The targets were examined by SEM and subjected to thermal shock test. The optimal thicknesses of target was substantiated, and the production yield for the 45Sc(p,n)45Ti reaction was calculated.

In the present study, ytterbium-169 was produced via the 169Tm(p, n)169Yb nuclear process at the AMIRS (Cyclone-30, IBA, Belgium) cyclotron, irradiating Tm2O3 with proton particles of 15 MeV primary energy and 20 μA current for 20 min. Deposition of Tm2O3 on Cu substrate was carried out via by the sedimentation method. The 543 mg of thulium(III)oxide with 108 mg of ethyl cellulose and 8 mL of acetone were used to prepare a Tm2O3 layer of 11.69 cm2. Yields of about 0.643 MBq 169Yb per μAh were experimentally obtained. 169Yb was separated in 80 ± 5% radiochemical yield using liquid–liquid extraction. Solvent extraction of no-carrier-added 169Yb from irradiated thulium(III)oxide target hydrochloric solution was investigated using di-(2-ethylhexyl)phosphoric acid (HDEHP).

101mRh is one of such radionuclides that has been considered as a potential candidate for targeted radiotherapeutic use, due to its nuclear decay and chemical properties. Electrodeposition of rhodium metal on a copper backing was performed in acidic sulphate. The target was bombarded with a current intensity of 120 μA (Ep = 29 → 25 MeV) for 30 min (60 μAh). Radiochemical methods were investigated to optimize the production of no-carrier-added 101Pd/101mRh. The use of a cyclotron target with radiochemical processes (i.e. electrodeposition, electrodissolution and ion-exchange column chromatography) were carried out to produce this radionuclide in high-specific activity and radiochemical form suitable for radiopharmaceutical syntheses.

Antimony-122, having a half-life of 2.723 d and I?-=97.59%, is an important radiotracer in studies of environmental contamination and food crops. For the work discussed in this paper, the production of 122Sb was done via the natSn(p, xn)122Sb nuclear reaction. Radiochemical separation was performed by silica gel column chromatography and liquid-liquid extraction methods. Excitation functions for the 122Sb radionuclide, via 122Sn(p, n)122Sb, natSn(p, xn)122Sb, 122Sn(d, 2n)122Sb, natSn(d, xn)122Sb, 124Sn(p, 3n)122Sb and 124Sn(d, 4n)122Sb reactions, were calculated by ALICE/91, ALICE/ASH, and TALYS-1.2 codes and compared with existing data.

Yttrium-88 (T1/2 = 106.6 d, Ib+ = 0.2% and IEC = 99.8%) is used in mixed gamma efficiency calibration standards and also as a substitute for 90Y to quantify the biodistribution of Y-pharmaceuticals in animals. Yttrium-88 and strontium-85 as gamma emitting radiotracers were produced via natSrCO3(p, xn) and natRbCl(p, xn) nuclear processes at AMIRS, with 18 and 15 MeV protons, respectively, at a current of 20 ?A for 10 hours. The deposition of natSrCO3 on the Cu backing was carried out by means of the sedimentation method. Yt- trium-88 was separated in a 92 ? 5% radiochemical yield using the precipitation technique by precipitating natSr as SrSO4 (strontium-85 as a tracer) while radioyttrium passed through the filter paper. Also, a theoretical study of the nuclear reaction cross-sections for proton and deuteron induced reactions on natSr and natRb for the production of 88Y and 85Sr, respectively, was performed using the EMPIRE (version 3.1 Rivoli), TALYS-1.26 codes and the TENDL-2010 database.

Aim of the presented study is to compare the calculated cross sections for the production of 140Nd/140Pr and 114mIn/114In via different reactions with incident particle energy up to 50 MeV as a part of systematic studies on particle-induced activations on metal target, theoretical calculation of production yield, calculation of target thickness requirement and suggestion for optimum reaction to produce Neodymium-140 and Indium-114m. 140Nd/140Pr and 114mIn/114In excitation functions via natCe(3He, xn)140Nd → 140Pr, 141Pr(p,2n)140Nd → 140Pr, 141Pr(d,3n)140Nd → 140Pr, 114Cd(p, n)114mIn → 114In, 114Cd(d,2n)114mIn → 114In and 116Cd(p,3n)114mIn → 114In reactions were calculated by ALICE/91, ALICE/ASH (GDH Model&Hybrid Model) and TALYS-1.2 (Equilibrium&Pre-Equilibrium) codes and compared to existing data.

This article presents, 122Sb (T 1/2 = 2.723 days, I β- = 97.59%) was produced via the natSn(p,xn) nuclear process at the AMIRS (Cyclone-30, IBA, Belgium). The electrodeposition experiments were carried out by potassium stannate trihydrate (K2Sn(OH)6) and potassium hydroxide. The optimum conditions of the electrodeposition of tin were as follows: 40 g/L natSn, 20 g/L KOH, 115 g/L K2Sn(OH)6, DC current density of 5 A/dm2 with a bath temperature of 75 °C. The electroplated Tin-target was irradiated with 26.5 MeV protons at current of 180 μA for 20 min. Solvent extraction of no-carrier-added 122Sb from irradiated Tin-natural target hydrochloric solution was investigated using di-n-butyl ether (C8H18O). Yields of about 3.61 MBq/μAh were experimentally obtained.

172Lu due to its suitable (T1/2 = 6.7 days) and high detection sensitivity, is used as a radiotracer in different fields. 172Lu appears to be suitable as a long-lived rare-earth tracer for compound labelling and biodistribution studies. In the present study, excitation functions via 172Yb(p,n)172Lu, natYb(p,xn)172Lu, 172Yb(d,2n)172Lu and natYb(d,xn)172Lu reactions were calculated by ALICE/91, ALICE/ASH and TALYS-1.2 codes. Deposition of natYb2O3 on Cu substrate was carried out via sedimentation method for the production of 172Lu. Cementation separation process and liquid–liquid extraction (LLX) of no-carrier-added (nca) radiolutetium from irradiated ytterbium(III)oxide target hydrochloric solution was described using Na(Hg) amalgam, α-hydroxyisobutyric acid (α-HIB) and di-(2-ethylhexyl)phosphoric acid (HDEHP).

Amongst the various radioisotopes of molybdenum, 93mMo in its no-carrier added form might be a potential candidate radionuclide in the field of nuclear medicine due to its suitable half-life (T1/2 = 6.85 h) and for its intense gamma line at 263.14 keV. Aim of the presented study is to compare the calculated cross sections for the production of 93mMo incident proton energy up to 30 MeV, theoretical calculation of production yield and calculation of target thickness requirement. Also, deposition of 93NbO2 on Cu substrate was carried out via two special sedimentation methods for the production of 93mMo and a simple and selective extraction of molybdenum was described.

Auger electron emitting radionuclides have potential for the therapy of small-size cancers because of their high level of cytotoxicity, low-energy, high linear energy transfer, and short range biologic effectiveness. Auger emitter 165Er (T1/2 = 10.3 h, IEC = 100%) is a potent nuclide for targeted radionuclide therapy. 165Er excitation function via 165Ho(p,n)165Er, 165Ho(d,2n)165Er, 166Er(p,2n)165Tm→165Er, 166Er(d,3n)165Tm→165Er, natEr(p,xn)165Tm→165Er and 164Er(d,n)165Tm→165Er reactions were calculated by ALICE/91, ALICE/ASH (GDH Model & Hybrid Model) and TALYS-1.2 (Equilibrium & Pre-Equilibrium) codes and compared to existing data. Requisite for optimal thicknesses of targets were obtained by SRIM code for each reaction.

Cancer is a leading cause of morbidity and mortality in the modern world. Treatment modalities comprise radiation therapy, surgery, chemotherapy and hormonal therapy. Radiation therapy can be performed by using external or internal radiation therapy. However, each method has its unique properties which undertakes special role in cancer treatment, this question is brought up that: For cancer treatment, whether external radiation therapy is more efficient or internal radiation therapy one? To answer this question, we need to consider principles and structure of individual methods. In this review, principles and application of each method are considered and finally these two methods are compared with each other.