Preparation of Rhenium-188-Lipiodol Using Freeze-Dried Kits for Transarterial Radioembolization: An Overview and Experience in a Hospital Radiopharmacy.

Background: Rhenium-188(188Re)-lipiodol is a clinically effective, economically viable radiopharmaceutical for Selective Internal Radiation Therapy of liver cancer. Present study evaluates the performance of three freeze-dried kits with respect to the radiochemistry, quality control, and overall "ease of preparation" aspects in a hospital radiopharmacy. Materials and Methods: Freeze-dried kits of acetylated 4-hexadecyl-4,7-diaza-1,10-decanedithiol (AHDD), super six sulfur (SSS), and diethyl dithiocarbamate (DEDC), obtained commercially or received as gift, were used for the preparation of 188Re-lipiodol using freshly eluted 188Re-sodium perrhenate from commercial Tungsten-188/188Re generator following recommended procedures. Results: The overall yield of 188Re-lipiodol prepared using AHDD Kit, SSS Kit, and DEDC Kit was 74.82% ± 3.3%, 87.55% ± 4.8%, and 76.38% ± 4.6%, respectively. Observed radiochemical purity (RCP) of 188Re-lipiodol prepared using these kits was 88.65% ± 2.8%, 92.92% ± 3.0%, and 91.38% ± 3.0%, respectively. Using a modified version of the DEDC Kits, overall yield of 87.17% ± 2.7% and RCP of 95.43% ± 2.3% could be achieved. Conclusions: While all three freeze-dried kits can be used for the preparation of 188Re-lipiodol in >70% overall yield, the modified version of DEDC Kits has some advantages in terms of preparation time and volume of Rhenium-188 activity that can be added to the kit vial. The latter feature of the DEDC Kit is particularly useful for patient dose preparation with 188Re activity of low radioactive concentration.

Radiochemical separation by liquid-liquid extraction for the determination of selenium in Mentha pulegium L.: Toxicity monitoring and health study.

In Algeria, Data and studies on the non-metal trace element selenium (Se) are presently lacking, therefore, the aim of this investigation is to provide new data on (Se) element via its determination for the first time from Mentha pulegium L. plant. The plant samples were collected in summer of 2012 from Ain-Oussera region, Djelfa province, Algeria; they were dried and powdered. After the neutron irradiation, the samples were digested using high oxidative reagents including H2SO4, HNO3, H2O2 and HCl. The end of this process gave two phases, organic and aqueous discard phase. By using a separating funnel, the organic phase was transferred into a vial in order to measure their induce radionuclide 75Se using gamma-ray spectrometer. A non-chromatographic and sensitive analytical technique RNAA (Radiochemical Neutron Activation Analysis), was applied in this investigation due to its great significant minor systematic error. Results were determined using two distinguish calculation methods, relative-RNAA and k0-RNAA, the findings were quite significant, whereas, the average separation yield was about 85% for both calculation methodologies. Moreover, (Se) concentration obtained from M. pulegium L., is close to the minimal FAO (Food and Agriculture Organization of the United Nations) recommended consumption.

Improved procedures of Sc(OH)3 precipitation and UTEVA extraction for 44Sc separation.

BACKGROUND: 44Sc is becoming attractive as a PET radionuclide due to its decay characteristics. It can be produced from 44Ca present in natural calcium with 2.08% abundance. MATERIALS AND METHODS: The targets were mostly prepared from natural CaCO3 or metallic calcium in the form of pellets. After irradiation they were dissolved in 3 M hydrochloric acid and 44Sc was separated from excess of calcium by precipitation of scandium hydroxide using ammonia. Alternatively, targets were dissolved in 11 M hydrochloric acid and 44Sc was separated by extraction chromatography on UTEVA resin. As the next step, in both processes 44Sc was further purified on a cation exchange resin. Initially, the separation procedures were developed with 46Sc as a tracer. Gamma spectrometry with a high purity germanium detector was used to determine the separation efficiency. Finally, the CaCO3 pellet with 99.2% enrichment in 44Ca was activated with protons via 44Ca(p,n)44Sc nuclear reaction. RESULTS: Altogether twenty two irradiations and separations were performed. The working procedures were developed and the quality of separated 44Sc solution was confirmed by radiolabeling of DOTATATE. The chemical purity of the product was sufficient for preclinical experiments. At the end of around 1 hour proton beam irradiation of CaCO3 pellet with 99.2% enrichment in 44Ca the obtained radioactivity of 44Sc was more than 4.8 GBq. CONCLUSION: 44Sc can be produced inexpensively with adequate yields and radionuclidic purity via 44Ca(p,n)44Sc nuclear reaction in small cyclotrons. The recovery yield in both investigated separation methods was comparable and amounted above 90%. The obtained 44Sc was pure in terms of radionuclide and chemical purity, as shown by the results of peptide radiolabeling.

Production of 15O for Medical Applications via the 16O(γ,n)15O Reaction.

15O (half-life, 122 s) is a useful radionuclide for PET applications. Current production of 15O typically makes use of the 14N(d,n)15O, 15N(p,n)15O, or 16O(p,pn)15O reactions using an accelerator. A novel approach for the production of 15O is via the 16O(γ,n)15O reaction using an electron linear accelerator. Photonuclear reactions using an electron linear accelerator may allow for feasible and economical production of 15O compared with the current methods. Methods: In this work, experiments using a repurposed Clinac were conducted using oxygen-containing alumina as a target material to study the production rate of 15O. Additional studies were conducted using a water target cell. Simulations using Geant4 were conducted to predict the activity and power dissipation in the target. Results: Bremsstrahlung radiation from the electron beam, and consequently 15O production via photonuclear reactions, is enhanced when a high-Z material, tungsten, is placed in front of the target. The alumina irradiations provided preliminary data to optimize the beam parameters and target configuration. The optimal thickness of tungsten was 1.4 mm for both the simulated and the measured studies of alumina. Simulations of irradiated water targets showed that tungsten thicker than 1.4 mm resulted in fewer photons available to activate the water; thus, a higher current was required to achieve a fixed dose. Alternatively, for a constant tungsten thickness, more power was deposited in the target with increasing beam energy, requiring a lower current to achieve a fixed dose. Actual irradiations of a water target yielded a quantity of 15O in the water that was consistent with expectations based on irradiations of alumina. Conclusion: Several parameters should be considered regarding the photonuclear production of 15O for an average patient dose of 1,850 MBq (50 mCi) in 10 mL. This work illustrates a variety of machine parameters capable of achieving a reasonable patient dose. Our simulations show that the power deposited in the target for these parameters is less than that in commercially operated cyclotron targets for the production of 18F. Thus, this work demonstrates that the photonuclear production of 15O may be a new production path for this useful radionuclide.

Studies on separation and purification of fission (99)Mo from neutron activated uranium aluminum alloy.

A new method has been developed for separation and purification of fission (99)Mo from neutron activated uranium-aluminum alloy. Alkali dissolution of the irradiated target (100mg) results in aluminum along with (99)Mo and a few fission products passing into solution, while most of the fission products, activation products and uranium remain undissolved. Subsequent purification steps involve precipitation of aluminum as Al(OH)3, iodine as AgI/AgIO3 and molybdenum as Mo-α-benzoin oxime. Ruthenium is separated by volatilization as RuO4 and final purification of (99)Mo was carried out using anion exchange method. The radiochemical yield of fission (99)Mo was found to be >80% and the purity of the product was in conformity with the international pharmacopoeia standards.

Tracer level radiochemistry to clinical dose preparation of (177)Lu-labeled cyclic RGD peptide dimer.

AIM: Integrin αvß3 plays a significant role in angiogenesis during tumor growth and metastasis, and is a receptor for the extracellular matrix proteins with the exposed arginine(R)-glycine(G)-aspartic acid(D) tripeptide sequence. The over-expression of integrin αvß3 during tumor growth and metastasis presents an interesting molecular target for both early detection and treatment of rapidly growing solid tumors. Considering the advantages of (177)Lu for targeted radiotherapy and enhanced tumor targeting capability of cyclic RGD peptide dimer, an attempt has been made to optimize the protocol for the preparation of clinical dose of (177)Lu labeled DOTA-E[c(RGDfK)]2 (E=Glutamic acid, f=phenyl alanine, K=lysine) as a potential agent for targeted tumor therapy. METHODS: (177)Lu was produced by thermal neutron bombardment on enriched Lu2O3 (82% in (176)Lu) target at a flux of 1 × 10(14) n/cm(2).s for 21 d. Therapeutic dose of (177)Lu-DOTA-E[c(RGDfK)]2 (7.4GBq) was prepared by adding the aqueous solution of the ligand and (177)LuCl3 to 0.1M NH4OAC buffer containing gentisic acid and incubating the reaction mixture at 90°C for 30 min. The yield and radiochemical purity of the complex was determined by HPLC technique. Parameters, such as, ligand-to-metal ratio, pH of the reaction mixture, incubation time and temperature were varied using tracer quantity of (177)Lu (37 MBq) in order to arrive at the optimized protocol for the preparation of clinical dose. Biological behavior of the radiotracer prepared was studied in C57/BL6 mice bearing melanoma tumors. RESULTS: (177)Lu was produced with a specific activity of 950 ± 50 GBq/mg (25.7 ± 1.4 Ci/mg) and radionuclidic purity of 99.98%. A careful optimization of several parameters showed that (177)Lu-DOTA-E[c(RGDfK)]2 could be prepared with adequately high radiochemical purity using a ligand-to-metal ratio ~2. Based on these studies therapeutic dose of the agent with 7.4 GBq of (177)Lu was formulated in ~63 GBq/µM specific activity with high yield (98.2 ± 0.7%), radiochemical purity and in vitro stability. Biodistribution studies carried out in C57/BL6 mice bearing melanoma tumors revealed specific accumulation of the radiolabeled conjugate in tumor (3.80 ± 0.55% ID/g at 30 min p.i.) with high tumor to blood and tumor to muscle ratios. However, the uptake of the radiotracer in the tumor was found to be reduced to 1.51 ± 0.32 %ID/g at 72 h p.i. CONCLUSIONS: The present work successfully demonstrates the formulation of an optimized protocol for the preparation of (177)Lu labeled DOTA-E[c(RGDfK)]2 for PRRT applications using (177)Lu produced by direct neutron activation in a medium flux research reactor.

Preparation of clinical-scale 99Mo/99mTc column generator using neutron activated low specific activity 99Mo and nanocrystalline γ-Al2O3 as column matrix.

INTRODUCTION: Preparation of clinical-scale (99)Mo/(99m)Tc generator using (n,γ) activated low specific activity (99)Mo and nanocrystalline γ-Al(2)O(3) as a high capacity sorbent matrix is attempted. METHODS: Nanocrystalline γ-Al(2)O(3) was synthesized by 'solid state mechanochemical' reaction of aluminum nitrate with ammonium bicarbonate. Experimental parameters were optimized to effectively separate (99m)Tc from (99)Mo using this sorbent as the column matrix. The performance features of a 13 GBq (350 mCi) (99)Mo/(99m)Tc generator using this sorbent and (99)Mo produced by (n,γ) route having specific activity 12.9-18.5 GBq/g were evaluated for 10 days. RESULTS: The sorbent possessed the requisite selectivity for (99)Mo and demonstrated a maximum sorption capacity of 200 ± 5mg Mo/g, which is ~10 times higher than that of ordinary acidic alumina. The overall yield of (99m)Tc was >80%, with radionuclidic purity >99.99% and radiochemical purity >99%. The yield of (99m)Tc varied from 7.8 to 2.1 GBq in the eluate for the six days of operation of the generator. The radioactive concentration of (99m)Tc eluted was adequate for the formulation of radiopharmaceuticals. The performance of the generator remained consistent over an extended period of 10 days. The eluted (99m)Tc was suitable for the formulation of (99m)Tc-DMSA and (99m)Tc-EC resulting in high radiolabeling yields (>98%). CONCLUSION: The effectiveness of γ-Al(2)O(3) as a new generation sorbent in the development of clinically useful (99)Mo/(99m)Tc generator using low specific activity (99)Mo and yielding (99m)Tc with adequate radioactive concentration and high purity suitable for formulation of radiopharmaceuticals is demonstrated.

Inter-laboratory comparison measurements of radiochemical laboratories in Slovakia.

The first inter-laboratory comparison organized by the radiochemistry laboratory of Water Research Institute (WRI) in Bratislava was carried out in 1993 and since then is it realized on an annual basis and about 10 radiochemical laboratories from all over Slovakia are participating. The gross alpha and gross beta activities, and the activity concentrations of (222)Rn, tritium, and (226)Ra, and U(nat) concentration in synthetic water samples are compared. The distributed samples are covering the concentration range prevailing in potable and surface waters and are prepared by dilution of certified reference materials. Over the course of the years 1993-2008, we observed the improvement in the quality of results for most of the laboratories. However, the success rate of the gross alpha determination activity is not improving as much as the other parameters.

Distribution of naturally occurring radionuclides (U, Th) in Timahdit black shale (Morocco).

Attention has been focused recently on the use of Moroccan black oil shale as the raw material for production of a new type of adsorbent and its application to U and Th removal from contaminated wastewaters. The purpose of the present work is to provide a better understanding of the composition and structure of this shale and to determine its natural content in uranium and thorium. A black shale collected from Timahdit (Morocco) was analyzed by powder X-ray diffraction and SEM techniques. It was found that calcite, dolomite, quartz and clays constitute the main composition of the inorganic matrix. Pyrite crystals are also present. A selective leaching procedure, followed by radiochemical purification and alpha-counting, was performed to assess the distribution of naturally occurring radionuclides. Leaching results indicate that 238U, 235U, 234U, 232Th, 230Th and 228Th have multiple modes of occurrence in the shale. U is interpreted to have been concentrated under anaerobic conditions. An integrated isotopic approach showed the preferential mobilization of uranium carried by humic acids to carbonate and apatite phases. Th is partitioned between silicate minerals and pyrite.

Biochemical assays for the characterization of DNA helicases.

Helicases are ubiquitous enzymes that disrupt complementary strands of duplex nucleic acid in a reaction dependent on nucleoside-5'-triphosphate hydrolysis. Helicases are implicated in the metabolism of DNA structures that are generated during replication, recombination, and DNA repair. Furthermore, an increasing number of helicases have been linked to genomic instability and human disease. With the growing interest in helicase mechanism and function, we have set out to describe some basic protocols for biochemical characterization of DNA helicases. Protocols for measuring ATP hydrolysis, DNA binding, and catalytic unwinding activity of DNA helicases are provided. Application of these procedures should enable the researcher to address fundamental questions regarding the biochemical properties of a given helicase, which would serve as a platform for further investigation of its molecular and cellular functions.

Comparison of inductively coupled plasma-mass spectrometry and radiochemical techniques for total uranium in environmental water samples.

Inductively coupled plasma-mass spectrometry (ICP-MS) method EPA 200.8 is gradually finding acceptance as an alternative to uranium analysis. A comparison of the ICP-MS with the accepted radiochemical method EPA 908.0 has been carried out based on data from laboratory control standards, national proficiency test samples, and environmental and drinking water samples from the State of Utah. The method detection limit (MDL) for ICP-MS was determined to be 0.017 microg/L or (0.011 pCi/L), and the minimum reporting limit (MRL) was 0.17 microg/L (MDL x 10) or (0.11 pCi/L). The minimum reporting limit for radiochemical 908.0 method is 1 pCi/L. Our spiked matrix recoveries, spiked blank samples, and reference materials deviate only a few percentage from the listed true values. Results demonstrate that the ICP-MS is a superior analytical tool for the determination of uranium in drinking and environmental waters at concentrations required by the United States Environmental Protection Agency.

A new method for separation of 131I, produced by irradiation of natural uranium.

Separation of 131I from natural uranium fission product mixtures has been accomplished by sorbing the 131I on special platinum-charcoal sorbent and desorbing by buffer solution (NaHCO3+Na2S2O3). High radiochemical and chemical purity is obtained by this method. Important parameters such as temperature, distillation rate, sorbing and desorbing rates and 131I separation yields have been investigated.

Determining the normal concentration of uranium in urine and application of the data to its biokinetics.

A method was developed to determine the concentration of uranium (238U) in urine. This method involved preconcentration of uranium from urine with calcium phosphate neutron activation followed by simple radiochemical separation. Using this method, it was possible to detect 0.2 ng of 238U. The average and geometric mean concentration of uranium found in urine of subjects from a normal background environment was 12.8 and 9.4 ng L-1, respectively. The concentration of uranium in the urine and the daily intake of uranium by the same population was used to arrive at the gastrointestinal absorption factor (f1) for uranium incorporated in the diet. The f1 factor was estimated to be 1.6%.