Production of iodine-124

Efficiency of 124I radioisotope production from natural and enriched tellurium dioxide using 124Te(p,xn)124I reaction

Background

¹²⁴I Iodine (T\documentclass[12pt]{minimal} \usepackage{amsmath} \usepackage{wasysym} \usepackage{amsfonts} \usepackage{amssymb} \usepackage{amsbsy} \usepackage{mathrsfs} \usepackage{upgreek} \setlength{\oddsidemargin}{-69pt} \begin{document}$$_{1/2}$$\end{document}1/2 = 4.18 d) is the only long-life positron emitter radioisotope of iodine that may be used for both imaging and therapy as well as for ¹³¹I dosimetry. Its physical characteristics permits taking advantages of the higher Positron Emission Tomography (PET) image quality, whereas the availability of new molecules to be targeted with ¹²⁴I makes it a novel innovative radiotracer probe for a specific molecular targeting.

Results

In this study Monte Carlo and SRIM/TRIM modelling was applied to predict the nuclear parameters of the ¹²⁴I production process in a small medical cyclotron IBA 18/9 Cyclone. The simulation production yields for ¹²⁴I and the polluting radioisotopes were  calculated for the natural and enriched ¹²⁴TeO₂  +  Al₂O₃  solid targets irradiated with 14.8 MeV protons. The proton beam was degraded energetically from 18 MeV with 0.2 mm Havar foil. The ¹²⁴Te(p,xn)¹²⁴I reactions were taken into account in the simulations. The optimal thickness of the target material was calculated using the SRIM/TRIM and Geant4 codes. The results of the simulations were compared with the experimental data obtained for the natural TeO₂ +Al₂O₃ target. The dry distillation technique of the 124-iodine was applied.

Conclusions

The experimental efficiency for the natural Te target was better than 41% with an average thick target (>0.8 mm) yield of 1.32 MBq/μAh. Joining the Monte Carlo and experimental approaches makes it possible to optimize the methodology for the ¹²⁴I production from the expensive Te enriched targets.

Radiochemistry, Production Processes, Labeling Methods, and ImmunoPET Imaging Pharmaceuticals of Iodine-124

Target-specific biomolecules, monoclonal antibodies (mAb), proteins, and protein fragments are known to have high specificity and affinity for receptors associated with tumors and other pathological conditions. However, the large biomolecules have relatively intermediate to long circulation half-lives (>day) and tumor localization times. Combining superior target specificity of mAbs and high sensitivity and resolution of the PET (Positron Emission Tomography) imaging technique has created a paradigm-shifting imaging modality, ImmunoPET. In addition to metallic PET radionuclides, 124I is an attractive radionuclide for radiolabeling of mAbs as potential immunoPET imaging pharmaceuticals due to its physical properties (decay characteristics and half-life), easy and routine production by cyclotrons, and well-established methodologies for radioiodination. The objective of this report is to provide a comprehensive review of the physical properties of iodine and iodine radionuclides, production processes of 124I, various 124I-labeling methodologies for large biomolecules, mAbs, and the development of 124I-labeled immunoPET imaging pharmaceuticals for various cancer targets in preclinical and clinical environments. A summary of several production processes, including 123Te(d,n)124I, 124Te(d,2n)124I, 121Sb(α,n)124I, 123Sb(α,3n)124I, 123Sb(3He,2n)124I, natSb(α, xn)124I, natSb(3He,n)124I reactions, a detailed overview of the 124Te(p,n)124I reaction (including target selection, preparation, processing, and recovery of 124I), and a fully automated process that can be scaled up for GMP (Good Manufacturing Practices) production of large quantities of 124I is provided. Direct, using inorganic and organic oxidizing agents and enzyme catalysis, and indirect, using prosthetic groups, 124I-labeling techniques have been discussed. Significant research has been conducted, in more than the last two decades, in the development of 124I-labeled immunoPET imaging pharmaceuticals for target-specific cancer detection. Details of preclinical and clinical evaluations of the potential 124I-labeled immunoPET imaging pharmaceuticals are described here.

Expanding PET-applications in life sciences with positron-emitters beyond fluorine-18

Positron-emission-tomography (PET) has become an indispensable diagnostic tool in modern nuclear medicine. Its outstanding molecular imaging features allow repetitive studies on one individual and with high sensitivity, though no interference. Rather few positron-emitters with near favourable physical properties, i.e. carbon-11 and fluorine-18, furnished most studies in the beginning, preferably if covalently bound as isotopic label of small molecules. With the advancement of PET-devices the scope of in vivo research in life sciences and especially that of medical applications expanded, and other than "standard" PET-nuclides received increasing significance, like the radiometals copper-64 and gallium-68. Especially during the last decades, positron-emitters of other chemical elements have gotten into the focus of interest, concomitant with the technical advancements in imaging and radionuclide production. With known nuclear imaging properties and main production methods of emerging positron-emitters their usefulness for medical application is promising and even proven for several ones already. Unfortunate decay properties could be corrected for, and β⁺-emitters, especially with a longer half-life, provided new possibilities for application where slower processes are of importance. Further on, (bio)chemical features of positron-emitters of other elements, among there many metals, not only expanded the field of classical clinical investigations, but also opened up new fields of application. Appropriately labelled peptides, proteins and nanoparticles lend itself as newer probes for PET-imaging, e.g. in theragnostic or PET/MR hybrid imaging. Furthermore, the potential of non-destructive in-vivo imaging with positron-emission-tomography directs the view on further areas of life sciences. Thus, exploiting the excellent methodology for basic research on molecular biochemical functions and processes is increasingly encouraged as well in areas outside of health, such as plant and environmental sciences.

Improved small scale production of iodine-124 for radiolabeling and clinical applications

AIM

This work describes a small-scale production of iodine-124 using a 16.5 MeV cyclotron, and a subsequent validation of the formulated sodium [124I]iodide solution for routinely clinical applications.

METHODS

Iodine-124 (¹²⁴I) was produced via the ¹²⁴Te(p, n)¹²⁴I reaction using a 16.5 MeV GE PETtrace® cyclotron. Irradiation was performed with a pre-prepared solid target consisting of [¹²⁴Te]TeO₂ (99.93%) and Al₂O₃. Different layer thicknesses, irradiation and extraction parameters were tested. After irradiation at the cyclotron, the shuttle with irradiated material was transferred fully automatically via a tube system to the Comecer ALCEO^® Halogen 2.0 extraction unit. Iodine-124 was subsequently extracted in form of sodium [¹²⁴I]iodide ([¹²⁴I]NaI) in 0.05 N aqueous NaOH solution, followed by reconstitution and validation for preclinical and clinical uses.

RESULTS

Good result was achieved using a beam degradation foil of 500 µm thickness in combination with beam currents between 10 and 15 µA. Under these conditions, up to 150 MBq no-carrier-added [¹²⁴I]NaI was obtained after a 2 h irradiation time in less than 500 µl 0.05 N NaOH. Isolation of [¹²⁴I]NaI, including evaporation and extraction at the ALCEO^® Halogen EVP unit was accomplished in 90 min 24 h after production (irradiation), the amount of iodine-123 as assessed by gamma-ray spectroscopy was less than 1.5%. The undesirable iodine-125 was not detectable by gamma spectroscopy. The extracted [¹²⁴I]NaI could be used directly for radiolabeling purposes, and after buffering with phosphate buffered saline (PBS) and sterile filtration for clinical applications.

CONCLUSIONS

Through the optimized conditions for irradiation and extraction, iodine-124 was produced in good radiochemical yields and high radionuclide purity. The generated injectable [¹²⁴I]NaI solution was sterile, non-pyrogenic and ready for preclinical and clinical applications after a sterile filtration through a 0.22 µm membrane filter.

Development of novel radionuclides for medical applications

Medical radionuclide production technology is well established. There is, however, a constant need for further development of radionuclides. The present efforts are mainly devoted to nonstandard positron emitters (eg, ⁶⁴ Cu, ⁸⁶ Y, ¹²⁴ I, and ⁷³ Se) and novel therapeutic radionuclides emitting low-range β^- particles (eg, ⁶⁷ Cu and ¹⁸⁶ Re), conversion or Auger electrons (eg, ^117m Sn and ⁷⁷ Br), and α particles (eg, ²²⁵ Ac). A brief account of various aspects of development work (ie, nuclear data, targetry, chemical processing, and quality control) is given. For each radionuclide under consideration, the status of technology for clinical scale production is discussed. The increasing need of intermediate-energy multiple-particle accelerating cyclotrons is pointed out.

An overview of (124)I production at a medical cyclotron by ALICE/ASH, EMPIRE-3.2.2 and TALYS-1.6 codes

Excitation functions of proton, deuteron and alpha induced nuclear reactions on enriched tellurium and antimony isotopes and also natural antimony were calculated by ALICE/ASH, EMPIRE-3.2.2 and TALYS-1.6 nuclear codes. Therefrom, the production yield of (124)I nuclide and its formed impurities were calculated by using the evaluation results. The calculated cross sections were compared with available experimental values in literatures. According to results, (124)Te(p,n)(124)I reaction is the method of choice due to formation of higher amount of (124)I nuclide and low levels of (125)I as an major concern in (124)I production.

124 Iodine: a longer-life positron emitter isotope-new opportunities in molecular imaging

(124)Iodine ((124)I) with its 4.2 d half-life is particularly attractive for in vivo detection and quantification of longer-term biological and physiological processes; the long half-life of (124)I is especially suited for prolonged time in vivo studies of high molecular weight compounds uptake. Numerous small molecules and larger compounds like proteins and antibodies have been successfully labeled with (124)I. Advances in radionuclide production allow the effective availability of sufficient quantities of (124)I on small biomedical cyclotrons for molecular imaging purposes. Radioiodination chemistry with (124)I relies on well-established radioiodine labeling methods, which consists mainly in nucleophilic and electrophilic substitution reactions. The physical characteristics of (124)I permit taking advantages of the higher PET image quality. The availability of new molecules that may be targeted with (124)I represents one of the more interesting reasons for the attention in nuclear medicine. We aim to discuss all iodine radioisotopes application focusing on (124)I, which seems to be the most promising for its half-life, radiation emissions, and stability, allowing several applications in oncological and nononcological fields.

Production of iodine-124 and its applications in nuclear medicine

Until recently, iodine-124 was not considered to be an attractive isotope for medical applications owing to its complex radioactive decay scheme, which includes several high-energy gamma rays. However, its unique chemical properties, and convenient half-life of 4.2 days indicated it would be only a matter of time for its frequent application to become a reality. The development of new medical imaging techniques, especially improvements in the technology of positron emission tomography (PET), such as the development of new detectors and signal processing electronics, has opened up new prospects for its application. With the increasing use of PET in medical oncology, pharmacokinetics, and drug metabolism, (124)I-labeled radiopharmaceuticals are now becoming one of the most useful tools for PET imaging, and owing to the convenient half-life of I-124, they can be used in PET scanners far away from the radionuclide production site. Thus far, the limited availability of this radionuclide has been an impediment to its wider application in clinical use. For example, sodium [(124)I]-iodide is potentially useful for diagnosis and dosimetry in thyroid disease and [(124)I]-M-iodobenzylguanidine ([(124)I]-MIBG) has enormous potential for use in cardiovascular imaging, diagnosis, and dosimetry of malignant diseases such as neuroblastoma, paraganglioma, pheochromocytoma, and carcinoids. However, despite that potential, both are still not widely used. This is a typical scenario of a rising new star among the new PET tracers.

Development of novel positron emitters for medical applications: nuclear and radiochemical aspects

In molecular imaging, the importance of novel longer lived positron emitters, also termed as non-standard or innovative PET radionuclides, has been constantly increasing, especially because they allow studies on slow metabolic processes and in some cases furnish the possibility of quantification of radiation dose in internal radiotherapy. Considerable efforts have been invested worldwide and about 25 positron emitters have been developed. Those efforts relate to interdisciplinary studies dealing with basic nuclear data, high current charged particle irradiation, efficient radiochemical separation and quality control of the desired radionuclide, and recovery of the enriched target material for reuse. In this review all those aspects are briefly discussed, with particular reference to three radionuclides, namely 64Cu, 124I and 86Y, which are presently in great demand. For each radionuclide several nuclear routes were investigated but the ( p,n) reaction on an enriched target isotope was found to be the best for use at a small-sized cyclotron. Some other positron emitting radionuclides, such as 55Co, 76Br, 89Zr, 82mRb, 94mTc, 120I, etc., were also produced via the low-energy (p,n), (p,α) or (d,n) reaction. On the other hand, the production of radionuclides 52Fe, 73Se, 83Sr, etc. using intermediate energy (p,xn) or (d,xn) reactions needs special consideration, the nuclear data and chemical processing methods being of key importance. In a few special cases, a high intensity 3He- or α-particle beam could be an added advantage. The production of some potentially interesting positron emitters via generator systems, for example 44Ti/44Sc, 72Se/72As and 140N d/140Pr is considered. The significance of new generation high power accelerators is briefly discussed.

The production of [124I]iodine and [86Y]yttrium

The use of paired tracers such as (124)I/(131)I and (86)Y/(90)Y allows pretherapy PET imaging with positron emitting radioisotopes of the same element as used for therapy. Whereas nowadays most therapy nuclides are produced by reactors or generators, the production of the corresponding PET isotopes requires the irradiation of adequate targets using particle accelerators such as cyclotrons. This paper describes the production routes for (124)I and (86)Y.

Fully automated production of iodine-124 using a vertical beam

A fully automated system for the production of iodine-124, based on techniques of vertical-beam irradiation and dry distillation, was developed. The system, coupled with a capsulated target, was able to irradiate the (124)TeO(2) target up to 29 μA for 1-4h, which yielded iodine-124 with an almost constant yield of 6.9 MBq/μAh at the end of bombardment. All procedures were performed automatically and repeatedly. The newly developed system would be suitable for routine, large-scale productions of iodine-124.

Iodine-124: a promising positron emitter for organic PET chemistry

The use of radiopharmaceuticals for molecular imaging of biochemical and physiological processes in vivo has evolved into an important diagnostic tool in modern nuclear medicine and medical research. Positron emission tomography (PET) is currently the most sophisticated molecular imaging methodology, mainly due to the unrivalled high sensitivity which allows for the studying of biochemistry in vivo on the molecular level. The most frequently used radionuclides for PET have relatively short half-lives (e.g. ¹¹C: 20.4 min; ¹⁸F: 109.8 min) which may limit both the synthesis procedures and the time frame of PET studies. Iodine-124 (¹²⁴I, t_1/2 = 4.2 d) is an alternative long-lived PET radionuclide attracting increasing interest for long term clinical and small animal PET studies. The present review gives a survey on the use of ¹²⁴I as promising PET radionuclide for molecular imaging. The first part describes the production of ¹²⁴I. The second part covers basic radiochemistry with ¹²⁴I focused on the synthesis of ¹²⁴I-labeled compounds for molecular imaging purposes. The review concludes with a summary and an outlook on the future prospective of using the long-lived positron emitter ¹²⁴I in the field of organic PET chemistry and molecular imaging.

Evaluation of excitation functions of proton and deuteron induced reactions on enriched tellurium isotopes with special relevance to the production of iodine-124

Cross-section data for the production of medically important radionuclide (124)I via five proton and deuteron induced reactions on enriched tellurium isotopes were evaluated. The nuclear model codes, STAPRE, EMPIRE and TALYS, were used for consistency checks of the experimental data. Recommended excitation functions were derived using a well-defined statistical procedure. Therefrom integral yields were calculated. The various production routes of (124)I were compared. Presently the (124)Te(p,n)(124)I reaction is the method of choice; however, the (125)Te(p,2n)(124)I reaction also appears to have great potential.

Production of no-carrier-added123Iviaheavy-ion activation of natural antimony oxide

Activation of natural Sb2O3with 48 MeV7Li3+beam results in the formation of no-carrier-added123I in the matrix along with the radionuclides123,125Xe and122Sb. The123I yield amounts to about 400 kBq/μAh and the radionuclidic impurity of124I to ∼1.2% of123I. Attempts to separate no-carrier-added iodine from bulk antimony target involved liquid-liquid extraction with TOA and HDEHP as well as precipitation of Sb2S3with thioacetamide. The precipitation technique was found to be the most effective for quantitative separation of123I from the bulk antimony oxide target.

Physicochemical and radiochemical aspects of separation of radioiodine from TeO2-targets

The radio-isotopes of iodine with mass-numbers 120, 123, and 124 are available via proton or deuteron induced reaction on enriched TeO2 targets with medical PET cyclotrons. The radioiodine can easily be separated from irradiated TeO2 targets using a dry thermo-chromatographic process. The half-time for the release of radioiodine from molten TeO2 samples as a function of the sample geometry and temperature has been studied systematically. Only 2 minutes are needed to release the radioiodine with >95% yield. Another 10 minutes are needed to trap the released iodine from the gas phase. An Al2O3 filter can be used to trap the released iodine intermediately and to clean the iodine from traces of evaporated TeO2. The activation energy for the release process has been determined to be QA = (170±30) kJ/mol which indicates, that convection contributes significantly to the release. For the adsorption-desorption process of iodine from Al2O3 the activation energy was measured to be Qad = (90±14) kJ/mol. The desorption half-time was <5 sec for T = 800°C. Dry Al2O3 columns can be used as a thermochromatographic 132Te-132I generator. The released iodine exhibits electrophilic properties, suitable for labelling procedures directly from the gas phase. Most likely the iodine is transported in the gas phase in the form of free radicals I*.

Standardized high current solid tellurium-124 target for cyclotron production of the radionuclides iodine-123,124

The production yield of iodine radionuclides (123/124I) by solid target irradiation strongly depends on the quality of the tellurium (Te) target. To avoid poor quality 124Te targets frequently obtained using the classical method, a new technology for high quality electroplated elemental 124Te was developed. In this method, simultaneous preparation of four smooth, dense and homogeneous 124Te targets was achieved within 1 h period. The prepared targets were capable of withstanding high beam currents without burning or losing the enriched material during bombardment. As a result, in comparison to the classical method, the overall yield of 123I increased by nearly 25%.

Thick tellurium electrodeposition on nickel-coated copper substrate for 124I production

Tellurium electrodeposition on a nickel-coated copper substrate was investigated for production of iodine-124. The electrodeposition experiments were carried out by the alkali plating baths. The optimum conditions of the electrodeposition of tellurium were as follows: 6 g l(-1) tellurium, pH=10, DC current density of ca. 8.55 mA cm(-2) and room temperature.

Optimization of 124I production via 124Te(p,n)124I reaction

(124)I was produced, via (124)Te(p,n)(124)I reaction, in greater than 3.7GBq (100 mCi, EOB) amount by bombarding (124)TeO(2) targets at 24 microA current for about 8h. This was achieved by keeping the target at 37 degrees relative to the beam during irradiation, by sweeping the beam across the target and by keeping the incident energy of the proton at 14.1MeV. The time-averaged yield of our 8h run was 21.1 MBq/microAh (0.57 mCi/microAh), which was 90% of the theoretical yield calculated using thick target yield data obtained from the reported excitation function for the reaction. At the end of bombardment, the level of (125)I and (126)I impurities, co-produced with (124)I, were 0.03% and 0.007%, respectively.

Alpha-particle induced reactions on natSb and 121Sb with particular reference to the production of the medically interesting radionuclide 124I

Excitation functions of the reactions (nat)Sb(alpha,xn)(123,124,125,126)I and (121)Sb(alpha,xn)(123,124)I were measured from their respective thresholds up to 26 MeV, with particular emphasis on data for the production of the medically important radionuclide (124)I. The conventional stacked-foil technique was used, and the samples for irradiation were prepared by a sedimentation process. The measured excitation curves were compared with the data available in the literature. From the experimental data the theoretical yields of the investigated radionuclides were calculated as a function of the alpha-particle energy. The calculated yield of (124)I from the (nat)Sb(alpha,xn)(124)I process over the energy range E(alpha) = 22-->13 MeV amounts to 1.02 MBq/microA x h and from the (121)Sb(alpha,n)(124)I reaction over the same energy range to 2.11 MBq/microA x h. The radionuclidic impurity levels are discussed. Use of (nat)Sb as target material would not lead to high-purity (124)I. Using highly enriched (121)Sb as target, production of (124)I of high radionuclidic purity is possible; the batch yield, however, is low.

Fluoro-, bromo-, and iodopaclitaxel derivatives: synthesis and biological evaluation

Paclitaxel (Taxol) is a clinically important chemotherapeutic agent. We describe the synthesis of fluoro-, bromo-, and iodopaclitaxel and their [(18)F]fluoro-, [(76)Br]bromo-, and [(124)I]iodo- analogues. [(18)F]Fluoropaclitaxel shows high uptake and rapid clearance from tissues in rats. Preadministration of paclitaxel in normal rats significantly increases (p < 0.005) retention of [(18)F]fluoropaclitaxel and [(76)Br]bromopaclitaxel in blood (33.0%), heart (32.0%), lung (37.6%) kidney (142.4%); and blood (33.4%), lung (42.3%), kidney (62.4%), respectively. [(18)F]Fluoropaclitaxel uptake in the brain of mdr1a/1b(-/-) mice is increased 1400% (p < 1.3e-07) relative to wild-type controls. Preadministration of paclitaxel or XR9576, a modulator, had little effect on the biodistribution in these mdr1a/1b(-/-) mice. As a result, [(18)F]fluoropaclitaxel will be a useful radiopharmaceutical for the study of multidrug resistant tumors.

Some optimisation studies relevant to the production of high-purity 124I and 120gI at a small-sized cyclotron

Optimisation experiments on the production of the positron emitting radionuclides 124I(T(1/2) = 4.18d) and (120g)I (T(1/2) = 1.35 h) were carried out. The TeO(2)-target technology and dry distillation method of radioiodine separation were used. The removal of radioiodine was studied as a function of time and the loss of TeO(2) from the target as a function of oven temperature and time of distillation. A distillation time of 15 min at 750 degrees C was found to be ideal. Using a very pure source and comparing the intensities of the annihilation and X-ray radiation, a value of 22.0 +/- 0.5% for the beta(+) branching in 124I was obtained. Production of 124I was done using 200 mg/cm(2) targets of 99.8% enriched 124TeO(2) on Pt-backing, 16 MeV proton beam intensities of 10 microA, and irradiation times of about 8 h. The average yield of 124I at EOB was 470 MBq(12.7 mCi). At the time of application (about 70 h after EOB) the radionuclidic impurity 123I (T(1/2) = 13.2 h) was <1%. The levels of other impurities were negligible (126I < 0.0001%;125I = 0.01%). Special care was taken to determine the 125I impurity. For the production of (120g)I only a thin 30 mg target (on 0.5 cm(2) area) of 99.9% enriched 120TeO(2) was available. Irradiations were done with 16 MeV protons for 80 min at beam currents of 7 microA. The 120gI yield achieved at EOB was 700 MBq(19 mCi), and the only impurity detected was the isomeric state 120 mI(T(1/2) = 53 min) at a level of 4.0%. The radiochemical purity of both 124I and 120gI was checked via HPLC and TLC. The radioiodine collected in 0.02 M NaOH solution existed >98% as iodide. The amount of inactive Te found in radioiodine was <1 microg. High purity 124I and 120gI can thus be advantageously produced on a medium scale using the low-energy (p,n) reaction at a small-sized cyclotron.

Excitation functions of 124Te(d,xn)124,125I reactions from threshold up to 14 MeV: comparative evaluation of nuclear routes for the production of 124I

Excitation functions of the nuclear reactions 124Te(d,xn)124-125I were measured from their respective thresholds up to 14.0 MeV via the stacked-foil technique. Thin samples were prepared by electrolytic deposition of 99.8% enriched 124Te on Ti-backing. The excitation function of the 124Te(d,n)125I reaction was measured for the first time. The present data for the 124Te(d,2n)124I reaction are by an order of magnitude higher than the literature experimental data but are in good agreement with the results of a hybrid model calculation. From the measured cross sections, integral yields of 124,125I were calculated. The energy range Ed = 14 --> 10 MeV appears to be the best compromise between 124I-yield and 1251-impurity. The calculated 124I-yield amounts to 17.5 MBq/microA h and the 125I-impurity to 1.7%. A critical evaluation of the three nuclear routes for the production of 124I, viz. 124Te(d,2n)-, 124Te(p,n)- and 125Te(p,2n)-processes, is given. The reaction studied in this work proved to be least suitable. The 124Te(p,n)-reaction gives 124I of the highest radionuclidic purity, and a small-sized cyclotron is adequate for production purposes. The 125Te(p,2n)-reaction is more suitable at a medium-sized cyclotron: the yield of 124I is four times higher than in the other two reactions but the level of 0.9% 125I-impurity is relatively high.

Excitation functions of 125Te(p, xn)-reactions from their respective thresholds up to 100 MeV with special reference to the production of 124I

Excitation functions of the nuclear reactions 125Te(p, xn) (119,120m, 120g, 121,122,123,124,125)I were measured for the first time from their respective thresholds up to 100 MeV using the stacked-foil technique. Thin samples were prepared by electrolytic deposition of 98.3% enriched 125Te on Ti-backing. In addition to experimental studies, excitation functions were calculated by the modified hybrid model code ALICE-IPPE. The experimental and theoretical data generally showed good agreement. From the measured cross section data, integral yields of (123,124,125)I were calculated. The energy range Ep 21 --> 15 MeV appears to be very suitable for the production of the medically interesting radionuclide 124I (T(1/2) = 4.18 d; I(beta)+ = 25%). The thick target yield of 124I amounts to 81 MBq/microA h and the level of 125I-impurity to 0.9%. The 125Te(p,2n)124I reaction gives 124I yield about four times higher than the commonly used 124Te(p,n)124I and 124Te(d,2n)124I reactions. The proposed production energy range is too high for small cyclotrons but large quantities of 124I can be produced with medium-sized commercial machines.

An efficient batch preparation of high specific activity

The increasing demand for radiolabeled metaiodobenzylguanidine (mIBG) prompted the need to obtain the radiopharmaceutical by a reliable, routine and simple synthetic method for batch production. The production of mIBG labeled with either 123I or 124I has been optimized by modifying literature methods that involve solid-state exchange reaction on "cold" mIBG facilitated by ammonium sulfate. The radiochemical yield and purity of radioiodinated mIBG generally exceeded 80 and 98%, respectively, with specific activity of > 50 mCi/mg.

Preparation of 124I solutions after thermodistillation of irradiated 124TeO2 targets

The 4.15-d radionuclide 124I is produced via the nuclear reaction 124Te(d, 2n) 124I by irradiation of 96% enriched 124TeO2 with 14 MeV deuterons, followed by thermodistillation. In order to minimise the loss of 124I, the quartz distillation tube was fitted to a stainless steel helix capillary trap directly behind the end of the furnace. Using this device, distillation yields of more than 80% were routinely obtained, and the activity was concentrated in markedly less than 100 microL solution. The 124I produced by this method proved to be useful for labelling proteins and IUdR.

Radiosynthesis and quality assurance of 5-[124I]Iodo-2'-deoxyuridine for functional PET imaging of cell proliferation

5-[124I]Iodo-2'-deoxyuridine ([124I]IUdR) was routinely produced by direct electrophilic labelling of 2'-deoxyuridine with 124I of high specific activity (12 Ci/micromol) in an Iodogen-coated ReactiVial, followed by purification on a Sep-Pak C-18 cartridge. The radiochemical purity was determined by TLC on a Silicagel-60 plate and by reverse-phase HPLC on a RP-18 column. Based upon 45 syntheses, the yield ranged from 45% to 65%. The radiochemical impurity of [124I]IUdR was determined at 2.9% by TLC (mainly iodate) and 4.3% by HPLC. The chemical stability of the solvated formulation allowed a time window of 2 days following end of synthesis (EOS) for chemical application, based upon the required 95% radiochemical purity grade of [124I]IUdR. The labelled compound was routinely used for the clinical determination of cell proliferation in glioma patients by positron emission tomography.

Radiation dose assessment in radioiodine therapy. Dose-response relationships in differentiated thyroid carcinoma using quantitative scanning and PET

Dose-response charts have been constructed to determine the tumouricidal dose for differentiated thyroid carcinoma metastases and thus enable precise activities of radioiodine to be prescribed in order to maximise tumour kill and minimise morbidity. Tumour and normal residual thyroid absorbed doses from radioiodine-131 have been determined with increased precision using a dual-headed whole-body rectilinear scanner with special high-resolution low-sensitivity collimators. Improved accuracy in the estimation of functioning tumour mass has been achieved using positron emission tomography (PET) with a low-cost large area PET camera. Dose-response data have been obtained for 33 patients. Following near-total thyroidectomy and 3.0 GBq 131I, a mean absorbed dose of 410 Gy achieved complete ablation of thyroid remnants in 75% of patients. Patients who had persistent uptake in the thyroid region on subsequent radioiodine scanning had received a mean dose of only 83 Gy. Cumulative absorbed doses in excess of 100 Gy were found to eradicate cervical node metastases. Patients with bone metastases, who generally have a poor prognosis, were found to have received doses of the order of only 20 Gy to the tumour deposits. The dose-response data explain the spectrum of clinical responses to fixed activities of radioiodine. In future, they will enable precise prescription of radioiodine to achieve tumouricidal doses whilst avoiding the morbidity and expense of ineffective therapy.

Treatment planning for 131I-mIBG radiotherapy of neural crest tumours using 124I-mIBG positron emission tomography

Patients designated to receive 131I-meta-iodobenzylguanadine (mIBG) for the treatment of neural crest tumours have been scanned with 124I-mIBG using the MUP-PET positron camera. Uptake was detected in tumour sites in lung, liver and abdomen. The tomographic images produced have allowed estimates to be made of the concentration of mIBG in both tumour and normal tissue. From these data it is possible to predict the radiation doses that would be achieved using therapy levels (up to 11 GBq) of 131I-mIBG. The levels of tumour uptake are between 0.5 and 2.0 kBq/g indicating that the radiation doses to tumour would be in the range 3 Gy to 7.5 Gy.

NCL-6-124I: a PET agent for the adrenal

The radiopharmaceutical 6 beta-[124I]iodomethyl-19-nor-cholest-5(10)-en-3 beta-ol (NCL-6-124I) was synthesized. The product was less sensitive to autoradiolytic decomposition in chloroform, than when stored as an injectable solution at 5 degrees C.