Development of [<Superscript>64</Superscript>Cu]-DOTA-anti-CD20 for targeted therapy

Copper-64 was produced as a by-product of ⁵⁵Co via ⁶⁴Ni(p,n)⁶⁴Cu by 15 MeV proton bombardment of ^natNi resulting in a thick target yield of 5.31 MBq/μAh (143.5 μCi/μAh) and a radiochemical separation yield of 95% (radionuclide purity >97% after 25 hours of bombardment). Rituximab was successively labeled with [⁶⁴Cu]-CuCl₂. N-succinimidyl-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA-NHS) was prepared at 25 °C using DOTA and N-hydroxy succinimide (NHS) in CH₂Cl₂ followed by the addition of 1 ml of a Rituximab pharmaceutical solution. Radiolabeling was performed at 37 °C in 3 hours. Radio thin-layer chromatography showed an overall radiochemical purity of 90–95% at optimized conditions (specific activity=30 GBq/mg, labeling efficacy; 82%) using various chromatography systems. The final isotonic ⁶⁴Cu-DOTA-Rituximab complex was passed through a 0.22 μm filter and checked by gel electrophoresis for radiolysis control. Stability of the final product was checked in the formulation and in presence of human serum at 37 °C.     

Accurate determination of half-life and radionuclidic purity of reactor produced <Superscript>177g</Superscript>Lu (<Superscript>177m</Superscript>Lu) for metabolic radiotherapy

Thanks to its favorable decay characteristics, ^177gLu is finding several applications in nuclear medicine, especially for palliative metabolic radiotherapy of cancer and radioimunotherapy. ^177gLu is produced in thermal nuclear reactor either by direct neutron capture ¹⁷⁶Lu(n,γ)^177(m+g)Lu on either natural or enriched ¹⁷⁶Lu target, or by reaction on enriched ¹⁷⁶Yb target followed by negatron decay. The latter method does produce a high radionuclidic purity and high specific activity radionuclide in no-carrier-added form, since ¹⁷⁷Yb decays solely to the ground state ^177gLu. Conversely, the first method does produce a low specific activity ^177gLu in carrier-added form,¹ contaminated by the long-lived radioisotopic impurity ^177mLu. The accurate determination of radionuclidic purity and half-life of ^177gLu carried out by HPGe and LSCS is presented in some details.     

Technology of DTPA and immunoglobulins conjugation and their attachment to <Superscript>90</Superscript>Y and <Superscript>177</Superscript>Lu radionuclides

The aim of the study of labeling of ligand–antibody conjugates was to find optimal conditions of preparing of these conjugates and appropriate radioactivity of selected nuclide for applications in nuclear medicine. Conjugation of the γ-immunoglobulin G (human or bovine IgG, polyclonal antibodies) and bifunctional chelating agent, diethylenetriaminepentaacetic acid dianhydride (cDTPAA), was carried out. Various values of the cDTPAA/antibody ratio, the weight concentration of polyclonal or monoclonal antibodies (MEM-97) and buffers were used. Further, the labeling conditions of the DTPA–IgG conjugate by radionuclides ⁹⁰Y and ¹⁷⁷Lu were optimized, and the labeling yield and the conjugation ratio of prepared radionuclide–DTPA–IgG conjugates was determined. Optimal incubation time of the immunoglobulin conjugation was obtained at 30 min from mixing of individual components. The labeling yield of radionuclide–DTPA–antibody conjugate higher than 95% was achieved. Higher values of conjugation ratio of radionuclide–DTPA–antibody conjugate were achieved in 0.1 mol L⁻¹ carbonate buffer, pH 8.5, and the 0.1 mol L⁻¹ carbonate buffer is suitable for studied conjugation systems. This study showed that the labeling yield as well as the conjugation ratio of tested systems depend on the amount of antibody substance, bifunctional chelating agent/antibody molar ratio and pH value of the buffer used.     

Radionuclide purity assessment and calibration of <Superscript>90</Superscript>Y(<Superscript>177</Superscript>Lu) glass particles using gamma-ray spectrometry

Intra-hepatic administration of radioactive glass microspheres is a treatment for patients with primary liver cancer and hepatic metastases. The purpose of this study was radionuclide purity assessment of new glass particles containing two radionuclide, ⁹⁰Y as a therapeutic source and also ¹⁷⁷Lu as a source of diagnostic gamma. For the mixed source, activity measurement using a dose calibrator cannot be used and we need new calibration methods. YAS (Yb) and YAS compositions were sol–gel derived glass particles and production of ⁹⁰Y (¹⁷⁷Lu) and ⁹⁰Y particles was performed using the Tehran Research Reactor. The radionuclide purity was carried out using γ-spectrometry with HPGe detector. A non-destructive spectroscopic assay was employed due to a newly updated low uncertainty positron branching ratio of ⁹⁰Y that emit 511 keV annihilation radiations. In another method, a new calibration of ⁹⁰Y using a non-destructive spectroscopic assay of ⁸⁸Y were investigated. Potential radionuclide impurity include: ⁸⁸Y, ¹⁵²Eu, ⁶⁰Co with activity 100, 50 and 5 Bq per 1 mg of that are not harmful for patients due to delivering radioactive particles about 20–50 mg in ⁹⁰Y(¹⁷⁷Lu) glass microspheres. Among of radionuclide impurity, ¹⁵²Er with a half life of 13.54 years and ⁸⁸Y with a half life of 106.65 days was important in the residual delivery device. For calibration of ⁹⁰Y with monitoring of 511 keV, errors were12.2–21%. In calibration of ⁹⁰Y using gamma spectroscopic assay of ⁸⁸Y, there was an error less than 14%. Spectroscopic assay of ⁸⁸Y can be performed easily and has more repeat for our purpose.     

<Superscript>234</Superscript>U/<Superscript>235</Superscript>U activity ratios as a probe for the <Superscript>238</Superscript>U/<Superscript>235</Superscript>U half-life ratio

High-resolution alpha-particle spectrometry was performed on three uranium materials enriched in ²³⁵U. Besides the ²³⁵U peaks, separate peaks belonging to impurity traces of ²³⁴U could be quantified. Relying on the isotopic composition of the uranium, as determined by mass spectrometry, the ratio of the half-lives of ²³⁸U and ²³⁵U was determined via the activity ratio of ²³⁴U and ²³⁵U in the materials. As an intermediate link, the ²³⁴U/²³⁸U half-life ratio was taken from published mass spectrometric analyses of ‘secular equilibrium’ uranium material. The resulting half-life ratio T _1/2(²³⁸U)/T _1/2(²³⁵U) = 6.351±0.031 is in agreement with the commonly adopted half-life values determined by Jaffey et al.     

Preparation of <Superscript>26</Superscript>Al, <Superscript>59</Superscript>Ni, <Superscript>44</Superscript>Ti, <Superscript>53</Superscript>Mn and <Superscript>60</Superscript>Fe from a proton irradiated copper beam dump

The station for pions cancer therapy was operated at PSI from 1980 to 1992. After a cooling time of 12 years it’s made of copper beam dump was cut and samples were taken for analytical purposes. The sampling collected about 500 g of high active copper chips that can be used for separation of exotic radionuclides. The analyses by gamma spectrometry, LSC and AMS showed main nuclides present to be ⁶⁰Co, ⁵⁴Mn, ²²Na, ⁶⁵Zn, ²⁶Al, ⁵³Mn, ⁵⁹Ni, ⁶³Ni, ⁵⁵Fe and ⁶⁰Fe and ⁴⁴Ti with a daughter nuclide ⁴⁴Sc. In the frame of ERAWAST project a procedure combining selective precipitation and ion exchange for the separation of the rare radionuclides from the copper beam dump was developed. The proposed separation procedure is easy for remote controlled implementation in a hot cell. The ion exchange separation of Ni, Al, Mg, Ti and Fe was complete and high decontamination factors for copper and cobalt were achieved. Based on the developed procedure a remotely controlled system for separation of exotic radionuclides from the copper chips was set up. The full scale system was installed in a hot cell where high activity levels can be handled. In order to evaluate the reliability and functionality of the system extensive tests have been done. During the test period 13.86 g in total of the proton irradiated copper beam dump were processed for separation of ²⁶Al, ⁵⁹Ni, ⁵³Mn, ⁴⁴Ti and ⁶⁰Fe. The results showed that the system was operational and the radionuclide separation was selective with high chemical yield. The procedure manages as well the generated liquid wastes containing high level of ⁶⁰Co activity.     

Sequential separation of <Superscript>99</Superscript>Tc, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe, <Superscript>90</Superscript>Sr and <Superscript>59/63</Superscript>Ni from radioactive wastes

A sequential separation procedure has been developed for the determination of ⁹⁹Tc, ⁹⁴Nb, ⁵⁵Fe, ⁹⁰Sr and ^59/63Ni in various radioactive wastes generated from nuclear power plants. Ion exchange and extraction chromatography were adopted for individual separation of the radionuclides. Precipitation was supplementarily utilized for both purification of the individual radionuclides and preparation of the radionuclide sources for use in a radioactivity measurement. The chromatographic separation behavior of the radionuclides both from the sample matrix metals and from one another was investigated using stable metals, Re (as a surrogate of ⁹⁹Tc), Nb, Fe, Sr and Ni. The validity of the procedure for reliability and applicability was evaluated by measuring the recovery of the metal carriers added to synthetic radioactive waste solutions. The recoveries by the chromatographic separation were in the range of 84.8 to 102.2% with 2s of less than 8.6%, the recoveries by the precipitation being in the range of 84.3 to 97.3% with 2s of less than 10.9%.     

Disequilibrium of dissolved <Superscript>234</Superscript>U/<Superscript>238</Superscript>U and <Superscript>210</Superscript>Po/<Superscript>210</Superscript>Pb in Greek rivers

For the first time, a radiological study for the dissolved ²³⁸U, ²³⁴U, ²¹⁰Pb and ²¹⁰Po was held in major Greek rivers across the country. ²³⁴U/²³⁸U activity ratios are above one in all samples and ²¹⁰Po/²¹⁰Pb activity ratios are respectively below the unit indicating the disequilibrium in the samples. Quite satisfactory correlations were observed among ²³⁴U and ²³⁸U as well as among ²¹⁰Po and ²¹⁰Pb values. Uranium isotopes were separated by ion exchange and electroplated on stainless steel plates. ²¹⁰Po was spontaneously deposited on nickel plates, while ²¹⁰Pb was indirectly determined through the ingrowth of ²¹⁰Po. The sources were measured by a-spectrometry.     

Systematic radiochemical separation for the determination of <Superscript>99</Superscript>Tc, <Superscript>90</Superscript>Sr, <Superscript>94</Superscript>Nb, <Superscript>55</Superscript>Fe and <Superscript>59,63</Superscript>Ni in low and intermediate radioactive waste samples

A simple and rapid separation procedure was systemized for the determination of ⁹⁹Tc, ⁹⁰Sr, ⁹⁴Nb, ⁵⁵Fe and ^59,63Ni in low and intermediate level radioactive wastes. The integrated procedure involves precipitation, anion exchange and extraction chromatography for the separation and purification of individual radionuclide from sample matrix elements and from other radionuclides. After separating Re (as a surrogate of ⁹⁹Tc) on an anion change resin column, Sr, Nb, Fe and Ni were sequentially separated as follows; Sr was separated as Sr (Ca-oxalate) co-precipitates from Nb, Fe and Ni followed by purification using Sr-Spec extraction chromatographic resin. Nb was separated from Fe and Ni by anion exchange chromatography. Fe was separated from Ni by anion exchange chromatography. Ni was separated as Ni-dimethylglyoxime precipitates after the removal of ^134,137Cs and ^110mAg by Cs-phosphotungstate and AgCl precipitation, respectively. Finally, the radionuclide sources were prepared by precipitation for their radioactivity measurements. The reliability of the procedure was evaluated by measuring the recovery of chemical carriers added to a synthetic radioactive waste solution.     

Determination of <Superscript>3</Superscript>H, <Superscript>226</Superscript>Ra, <Superscript>222</Superscript>Rn and <Superscript>238</Superscript>U in Austrian ground- and drinking water

Screening measurements for ³H, ²²⁶Ra, ²²²Rn and ²³⁸U in ground water were performed within a ground- and drinking water project in Austria. The aim of this project is to get an overview of the distribution of natural radionuclide activity concentration levels in ground water bodies. In some cases this water is used for drinking water abstraction. In this paper methods and results of the screening measurements are presented. Regions with high activity concentrations were identified and in these regions further investigation for ²²⁸Ra, ²¹⁰Pb and ²¹⁰Po will be conducted.     

Extraction chromatography for the separation of <Superscript>239</Superscript>Np from <Superscript>243</Superscript>Am

Summary An extraction chromatography method was developed for the separation of 239Np from ²⁴³Am in nitric acid solution. A sorbent based on aliphatic quaternary amine Aliquat-336 and hydrophobized silica gel was prepared. ²³⁹Np reduced to the oxidation state(IV) with ferrous sulfamate in 2M or 6M HNO₃ sorbs on the prepared silica gel column. After washing with 0.1M ferrous sulfamate in 2.5M HNO₃, ²³⁹Np is eluted with 0.1M HNO₃ containing 0.02M HF. The separation of ²⁴³Am from ²³⁹Np is very effective. The purity of ²³⁹Np was found to be better than 99.5%. The proposed ²³⁹Np milking procedure is suitable for the preparation of ²³⁹Np tracer that can be used for the determination of ²³⁷Np radiochemical yield.     

Measurements of <Superscript>210</Superscript>Po and <Superscript>210</Superscript>Pb in total diet samples: Estimate of dietary intakes of <Superscript>210</Superscript>Po and <Superscript>210</Superscript>Pb for Japanese

To estimate the dietary intakes of ²¹⁰Pb and ²¹⁰Po for the Japanese adults and their annual effective doses, ²¹⁰Pb and ²¹⁰Po were measured for 240 daily diet samples collected at two locations of Ishikawa Prefecture in Japan over three years by duplicate portion studies. No appreciable differences in intake rates of ²¹⁰Pb and ²¹⁰Po and their ²¹⁰Po/²¹⁰Pb ratios were seen among the years in each district, and between the two districts. The intake rates evaluated using 240 diet samples were 0.20 Bq/d/p for ²¹⁰Pb and 0.61 Bq/d/p for ²¹⁰Po as a median, respectively. Annual effective doses of ²¹⁰Pb and ²¹⁰Po for Japanese adults were estimated to be 0.050 and 0.053 mSv/y, respectively.     

Distribution of <Superscript>14</Superscript>C, <Superscript>90</Superscript>Sr and <Superscript>228</Superscript>Th in an elephant tusk

The period of date of death of an elephant can be assessed by analyzing four different radionuclides, ¹⁴C, ⁹⁰Sr, ²²⁸Th and ²³²Th in its ivory. These nuclides are supposed to have variing concentrations at different parts of a tusk. The reason is the procedure of growth which takes place at the butt-site of a tusk. Therefore the site of sampling could have a big influence on the assessed date of death. However, to find out if the position of sampling is important a complete tusk was analyzed regarding the distribution of these nuclides. Results show that the concentration activity of ¹⁴C and ²²⁸Th varies in different parts of a tusk. The activity concentration of ⁹⁰Sr is very similar in all analyzed parts. The conclusion is that sampling at the butt of a tusk is recommended for age assessment.     

Determination of <Superscript>241</Superscript>Am/<Superscript>243</Superscript>Am ratios

Determination of ²⁴¹Am/²⁴³Am ratios is required for vanous purposes including assay of Am by isotope dilution techniques. Alpha-spectrometry on electrodeposited sources is a preferred technique for this determination. However, there is an inherent problem of tail contribution which necessitates the use of suitable algorithms to account for the same. Recently, in the frame of a Coordinated Research Program (CRP) of the International Atomic Energy Agency (IAEA), WinALPHA software has been developed which is a combination of an asymmetrical Gaussian for the main part of the peak and a low energy function. Therefore, it was of interest to compare the use of this algorithm with the routinely used method, in our laboratory, based on geometric progression (G. P.) decrease. Since, there are no reference materials available commercially for ²⁴¹Am/²⁴³Am ratios, synthetic mixtures covening a wide range (0.3 to 2.0) of ²⁴¹Am/²⁴³Am α-activity ratios were used and un-ignited electrodeposited sources were prepared for α-spectrometry. The α-spectra obtained using PIPS detector, were evaluated using the two algonthms The ²⁴¹Am/²⁴³Am α-activity ratios obtained were also compared with those determined by thermal ionization mass spectrometry (TIMS). An agreement of about 1% was obtained in the ²⁴¹Am/²⁴³Am ratios determined by the two methods and also by using the two algorithms for α-spectrum evaluation.     

Mixing of atmospheric <Superscript>210</Superscript>Pb and <Superscript>7</Superscript>Be and <Superscript>137</Superscript>Cs and <Superscript>90</Superscript>Sr fission products in four characteristic soil types

Depth distribution of atmospheric ²¹⁰Pb and ⁷Be and ⁹⁰Sr and ¹³⁷Cs fission products was measured in two types of aeolian soils (desert dust and volcanic ash), irrigated paddy soil and strongly acidic soil. The depth dependence of ²¹⁰Pb, ⁷Be and ¹³⁷Cs show that these radionuclides have been diffused as solid soil particles in surface soil layers. In aeolian soil layers, about 50% of ⁹⁰Sr were diffused in surface soil layer and the remaining 50% had penetrated to deeper layers. The half of the fission particles containing ⁹⁰Sr were shown to have decomposed over the past 35 years.     

Production of <Superscript>41</Superscript>Ar and <Superscript>79</Superscript>Kr gaseous radiotracers for industrial applications

Radiotracers are extensively used in many industries for trouble shooting and optimization of process parameters leading to considerable savings in time and huge economic benefits. In chemical and petrochemical industries different gases and vapours flowing in the conversion reactors play a major role in the final production. Gaseous radiotracers are ideal to study hydrodynamics of gas phases in process vessels. ⁴¹Ar and ⁷⁹Kr are the preferred gaseous radiotracers for such studies. Owing to the increase in demand from Indian industries for gas phase radiotracers, efforts have been made to produce ⁴¹Ar and ⁷⁹Kr indigenously by irradiation of ⁴⁰Ar and enriched ⁷⁸Kr gaseous targets in research reactors. Prequalification of the containers used, safety aspects concerning accidental rupture and mandatory tests necessary for irradiation of gaseous targets in the reactors have been studied. The paper describes some of the important safety aspects involved and the results of trial irradiations on the production of ⁴¹Ar and ⁷⁹Kr radiotracers. Standardization of suitable assay protocols for their regular production and supply for applications in industries is also described.