A novel electrochemical <Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generator

A novel electrochemical process to avail clinical grade ^99mTc from (n,γ)⁹⁹Mo has been demonstrated. The electrochemical parameters were optimized to maximize the ^99mTc yield with minimal ⁹⁹Mo contamination. ⁹⁹Mo/^99mTc generators containing up to 29.6 GBq (800 mCi) ⁹⁹Mo were developed and their performance were extensively evaluated for 10 days without changing the operating conditions. Very high radioactive concentration of ^99mTcO₄ ⁻ of acceptable quality, commensurate with hospital radiopharmacy requirements could be availed from the system with >90% yield. The compatibility of the product for the formulation of ^99mTc labeled radiopharmaceuticals such as ^99mTc-DMSA and ^99mTc-EC was found to be satisfactory in terms of high labeling yields. The proposed route represents an important step for enhancing the scope of accessing clinical grade ^99mTc from low specific activity (n, γ)⁹⁹Mo.     

Radiostrontium separation from sodium molybdate solution and its measurement using LSA: an application to radiopharmaceutical analysis

Technetium (^99mTc), a decay product of molybdenum (⁹⁹Mo), is employed as radioisotope in nuclear medicine. Several practical devices known as generators are commercially available which enable the user to separate the daughter from the parent radionuclide. The present study is focused on quality control of chromatographic technetium generator. A properly constructed generator should comply with international requirements of radionuclide purity of ⁹⁰Sr/⁹⁹Mo ≤ 6 × 10^?8 and ⁸⁹Sr/⁹⁹Mo ≤ 6 × 10^?7. For this purpose an analytical method was optimized to quantify radiostrontium (⁸⁹Sr and ⁹⁰Sr) in sodium molybdate [Na ₂ ⁹⁹ MoO₄] solution, a fission product used for ⁹⁹Mo/^99mTc generators. Dowex 1 × 8 and alumina were used in sequence followed by tributyl phosphate extraction for radiostrontium separation. Cerenkov measurement of ⁸⁹Sr and ⁹⁰Sr (through its descendent ⁹⁰Y) was performed using Perkin Elmer Tricarb LSA 3170 with detection efficiency of 42 and 14 %, respectively. Since efficiency of Cerenkov counting is sensitive to presence of color, spectral index of sample was used to correct the counting efficiency. The chemical recovery for strontium was 22 % and for yttrium was 80 % as determined by inductively coupled plasma optical emission spectrometry. Lower limit of detection was found to be 6.3 and 14.4 Bq L^?1 for ⁹⁰Sr and ⁸⁹Sr, respectively with 60 min counting time. Hence method can be applied successfully to analyze ^89,90Sr in fission molybdenum used as radiopharmaceutical with a relative error of <10 %.     

An alternative route for the preparation of the medical isotope 99Mo from the 238U(γ, f) and 100Mo(γ, n) reactions

The radionuclide ⁹⁹Mo, which has a half-life of 65.94 h was produced from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions using a 10 MeV electron linac at EBC, Kharghar Navi-Mumbai, India. This has been investigated since the daughter product ^99mTc is very important from a medical point of view and can be produced in a generator from the parent ⁹⁹Mo. The activity of ⁹⁹Mo was analyzed by a γ-ray spectrometric technique using a HPGe detector. From the detected γ-rays activity of 140.5 and 739.8 keV, the amount of ⁹⁹Mo produced was determined. For comparison, the amount of ⁹⁹Mo from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions was also estimated using the experimental photon flux from ¹⁹⁷Au(γ, n)¹⁹⁶Au reaction. The amount of ⁹⁹Mo from the detected γ-lines is in agreement with the estimated value for ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions. The production of ⁹⁹Mo activity from ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions is a relevant and novel approach, which provides alternative routes to ^235,238U(n, f) and ⁹⁸Mo(n, γ) reactions, circumventing the need for a reactor. The viability and practicality of the ⁹⁹Mo production from the ²³⁸U(γ, f) and ¹⁰⁰Mo(γ, n) reactions alternative to ^235,238U(n, f) and ⁹⁸Mo(n, γ) reactions has been emphasize. An estimate has been also arrived based on the experimental data of present work to fulfill the requirement of DOE.     

Feasibility study for production of 99mTc by neutron irradiation of MoO3 in a 250 kW TRIGA Mark II reactor

The subject of this paper is to explore the possibility to obtain ^99mTc from activation of ⁹⁸Mo, using the TRIGA Mark II low flux research reactor (Vienna, Austria). Irradiation of both natural and enriched in ⁹⁸Mo molybdenum oxides was compared. Aims of this work included the determination of neutron fluxes and ⁹⁸Mo(n, γ)⁹⁹Mo reaction effective cross section in the TRIGA Mark II reactor irradiation channels, calculation of ⁹⁹Mo specific activities, determination of optimal irradiation conditions for the subsequent ^99mTc separation from MoO₃ targets using concentrating technologies.     

Selective separation of99Mo from fission products in chloride media on activated alumina

The radioisotope⁹⁹Mo generated through the fission of neutron-irradiated uranium targets was separated in an extremely radiochemically pure state. The target dissolution mixture was run on a silica gel column whereby the⁹⁵Zr−⁹⁵Nb activity, the serious radiocontaminants of⁹⁹Mo, was adsorbed on the gel. The effluent from the gel column was processed and treated with sodium dihydrogen phosphate. The processed solution was run on an activated alumina column whereby⁹⁹Mo was adsorbed as phospho-molybdate complex ion, then desorbed from the column as the molybdate ion by eluting with 10% NaOH solution. By recycling the adsorption and desorption steps on alumnia,⁹⁹Mo was finally obtained totally free from all other radiocontaminants.     

The Intermediate Level Liquid Molybdenum-99 Waste Treatment Process at the Australian Nuclear Science and Technology Organisation

ANSTO manufactures ⁹⁹Mo for radiopharmaceutical use. Alkaline Intermediate Level Liquid Waste (ILLW) from this process, plus legacy acidic waste, are planned to be treated by converting both wastes into stable, solid, waste forms with oxide-basis loadings of 25-35 wt% and 30-50 wt%, respectively. The hot-cell plant design utilises the same unit process steps to treat both wastes. Hot-Isostatic Pressing (HIP) is employed to consolidate the processed waste and achieve substantial waste volume reductions compared to a cementation option. In this paper an overview of the treatment process and selected waste forms for ANSTO's ⁹⁹Mo production ILLW is given.     

Radiochemical separation of99Mo from natural uranium irradiated with thermal neutrons

⁹⁹Mo was separated from uranium and insoluble fission product hydroxides. More than 98% of⁹⁹Mo radioactivity was extracted with bis (2-ethylhexyl)phosphoric acid. The organic phase was washed and⁹⁹Mo was back-extracted from the organic phase with NH₄OH solution. The percent recovery from the organic phase was 91% and the purity of⁹⁹Mo was more than 99%. Pure^99mTc was also extracted from the organic phase with a saline solution. Reversed-phase partition chromatography was used for the purification of⁹⁹Mo from¹³¹I and other fission products (10% HDEHP on kieselguhr bed).¹³¹I and other isotopes were quantitatively eluted with 0.1M H₂SO₄,⁹⁹Mo was eluted using a mixture of 0.5 M HCl and 30% H₂O₂.     

Solid targets for 99mTc production on medical cyclotrons

Recent disruptions in the molybdenum-technetium generator supply chain prompted a review of non-reactor based production methods for both ⁹⁹Mo and ^99mTc. Small medical cyclotrons (E _p ~ 16–24 MeV) are capable of producing Curie quantities of ^99mTc from isotopically enriched ¹⁰⁰Mo using the ¹⁰⁰Mo(p,2n)^99mTc reaction. Unlike most other metallic target materials for routine production of medical radioisotopes, molybdenum cannot be deposited by reductive electroplating from aqueous salt solutions. To overcome this issue, we developed a new process for solid molybdenum targets based on the electrophoretic deposition of fine ¹⁰⁰Mo powder onto a tantalum plate, followed by high temperature sintering. The targets obtained were mechanically robust and thermally stable when irradiated with protons at high power density.     

The quality of99mTc eluates

The possible effects of several protecting procedures on the quality of^99mTc eluates were investigated. The content of⁹⁹Mo in the eluates (⁹⁹Mo breakthrough) was expressed in (%) with respect to the total adsorbed⁹⁹Mo radioactivity and in () i.e. as the ratio of⁹⁹Mo and^99mTc radioactivities in each particular eluate. The radiochemical purity was expressed in (%) of^99mTc(VII) in the eluates. The content of Al³⁺ and Cu²⁺ as chemical impurities was also determined.     

99mTc-glucoheptonate-guanine: Synthesis, biodistribution and imaging in animals

The aim of the current study was to design a nucleotide-based radiopharmaceutical which could be labeled with ^99mTc and to investigate its radiopharmaceutical efficiency and stability. GHA (glucoheptonate) was used as bifunctional chelate. GHA was labeled with ^99mTc by SnCl₂ reduction method first, and then G (guanine) was conjugated with ^99mTc-GHA at 90 °C. In order to determine its radiopharmaceutical stability, thin layer radio chromatography (TLRC) and electrophoresis were employed. In addition, the results were confirmed using high performance liquid radio chromatography (HPLRC). Scintigraphic imaging was performed on rats with mammary tumors, while tissue distribution was determined on Albino Wistar rats. Labeling yield was found to be over 95% and the labeled complex maintained its stability during the study period. The lipophilicity of the ^99mTc-GHG was measured and the partition coefficient (logP) of the labeled compound calculated. The results demonstrated that the uptake of ^99mTc-GHG (^99mTc-glucoheptonate-guanine) reached its maximum at 3 hours p.i. in stomach and intestines. Main way of excretion was renal. Hepatobiliary excretion was also observed. In conclusion, ^99mTc-GHG may be useful as a nucleotide-based radiopharmaceutical for in vivo applications.     

Evaluation of proton induced reactions on 100Mo: New cross sections for production of 99 mTc and 99Mo

The use of the ⁹⁹Mo^99mTc generator in nuclear medicine is well established world wide. The production of the ⁹⁹Mo (T_1/2 = 66 h) parent as a fission product of ²³⁵U is largely based on the use of reactor technology. From the early 1990's accelerator based production methods to provide either direct produced ^99mTc or the parent ⁹⁹Mo, were studied and suggested as potential alternatives to the reactor based production of ⁹⁹Mo. A possible pathway for the charged particle production of ^99mTc and ⁹⁹Mo is irradiation of molybdenum metal with protons via the reaction ¹⁰⁰Mo(p,2n)^99mTc and ¹⁰⁰Mo(p,pn)⁹⁹Mo, respectively. The earlier published excitation functions show large differences in their maximum that result in large differences in the calculated yields. We therefore decided to study the excitation function for these proton-induced reactions. In this work the newly measured excitation functions as well as an evaluation of earlier measured data and a discussion of the observed disagreements are presented.     

Issues with the European Pharmacopoeia Quality Control Method for 99mTc-Labelled Macroaggregated Albumin

Technetium-99m macroaggregated albumin ([^99mTc]Tc-MAA) is an injectable radiopharmaceutical used in nuclear medicine for lung perfusion scintigraphy. After changing to a new batch of macroaggregated albumin (MAA), we saw unwanted uptake in the liver and spleen. The batch was therefore tested by both the supplier and us and we found it to comply with the requirements of the European Pharmacopoeia (Ph. Eur.). However, a simple comparison between the problematic batch and a batch supplied by another manufacturer showed that there was a significant difference. The quality testing showed a higher number of small particles in the problem encumbered MAA batch with unwanted in vivo uptake. In this article we present a simple method of testing for particle size of [^99mTc]Tc-MAA, which gives a good indication of how the radioactive drug performs in vivo. We argue that the quality control method described in the Ph. Eur. should be changed. The changes will improve concordance between the laboratory analyzes and what is seen in vivo in human lung perfusion scintigraphy. Furthermore, we hope that the MAA suppliers without delay will replace their release procedure to be in accordance with the method described in this article.     

<Superscript>99</Superscript>Mo/<Superscript>99m</Superscript>Tc generators based on aluminum molybdate gel matrix prepared by nano method

A procedure for preparation of ⁹⁹Mo/^99mTc radioisotope generator based on low specific activity neutron activated ⁹⁹Mo was developed. Aluminum molybdate(VI)-⁹⁹Mo of high Mo(VI) content (~?364 mg/g Al⁹⁹Mo) was prepared by mixing low specific activity molybdate(VI)-⁹⁹Mo and aluminum mixture solution with isoamyl alcohol. Al⁹⁹Mo gel matrix was precipitated when the pH of the mixture solution was raised to ~?5 by addition of NaOH to the mixture. Radiometric measurements indicate the strong fixation of Molybdate(VI)-⁹⁹Mo species in the form of the sparingly insoluble Al⁹⁹Mo gel matrix. The prepared AlMo gel matrix was physiochemically characterized. Al⁹⁹Mo gel matrix was used as a base material for preparation of ⁹⁹Mo/^99mTc generator. The ^99mTc eluted from ⁹⁹Mo/^99mTc radioisotope generator was found to have relatively high elution yield (84?±?2.3%), radionuclidic (≥?99.99%), radiochemical (98.1?±?0.9%) and chemical purity.     

Elution of 235U fission products through a calcined hydrotalcite-packed column

Calcined hydrotalcite packed columns were utilized to sorb ²³⁵U fission products and their decay products. The elution behavior of some radionuclides was studied after washing the columns, either with distilled water or 0.5% NaCl solution. Afterwards, fission products and their decay products were eluted using 0.5% NaCl solution. It was found that no matter the washing process, ^99mTc, the b^--decay product of ⁹⁹Mo, was easily separated from ⁹⁹Mo which was strongly retained on the hydrotalcite. ¹³²I, the b^--decay product of ¹³²Te, was eluted slowly and was separated from ¹³²Te which was retained on the column. ¹³¹I and ¹⁴⁰Ba were eluted together with ^99mTc and ¹³²I, although in smaller proportions.     

Pharmaceutical grade sodium [99mTc] pertechnetate from low specific activity 99Mo using an automated 99Mo/99mTc-TCM-autosolex generator

Performance study of a computer controlled automated closed cyclic module for the separation and recovery of ^99mTc from low specific activity (n, γ) ⁹⁹Mo using methyl ethyl ketone (MEK) solvent extraction technique named ⁹⁹Mo/^99mTc-TCM-AUTOSOLEX (Technetium automated solvent extraction) Generator is described. The entire system is automated and controlled by a user-friendly PC based graphical user interface that actually supervises process via an embedded system based electronic controller. The average yield of separation of ^99mTc was above 85 % and ⁹⁹Mo breakthrough in ^99mTc pertechnetate was <0.002 %. The sodium pertechnetate obtained was a clear solution having pH 6–7, Radiochemical (RC). Purity >99 %, MEK content <0.1 % (v/v), Al and Mo content <10 µg/ml. R. C. Purity of ^99mTc-radiopharmaceuticals studied was not less than 96 %. Bio-Quality control studies confirm that sodium pertechnetate obtained was sterile and pyrogen free. Imaging studies in animals and humans with limited radiopharmaceuticals show that the quality of ^99mTc-pertechenate obtained in the present module was good enough to do clinical study.     

Application of neutron activation analysis in a fission molybdenum separation study

The separation of⁹⁹Mo from low-enriched uranium (LEU, 19.5%²³⁵U) targets was evaluated using natural uranium (NU) and non-radioactive tracers. Neutron activation analysis was used to determine (1) the efficiency of molybdenum recovery and (2) the decontamination factor of numerous fission product elements from the molybdenum product. Using NU and non-radioactive elements simplified procedures and allowed tests to be completed in a fume hood instead of a shielded cell. During activation of the non-radioactive tracers, uranium fission occurs, which can interfere with subsequent gamma-ray analysis. A comparison was made of the interferences caused by these fission products from both NU and LEU.     

Radiobiologic /tissue/ profiles of99mTc-diagnostics: behaviour of99mTc-N/p-butylphenylcarbamoylmethyl/iminodiacetic acid, /99mTc-p-butyl HIDA/, in normal rats

The biologic profiles of a hepatobiliary radiodiagnostic agent,^99mTc-N/p-butylphenylcarbamoylmethyl/iminodiacetic acid,^99mTc-p-butyl HIDA for short, are described in terms of its pharmacokinetics in normal rats over a 24 h period. It could be used to monitor the quality of this hepatobiliary radiopharmaceutical if prepared in an in-house hospital radiopharmacy.     

Sensitivity study on modeling radioxenon signals from radiopharmaceutical production facilities

As part of the Comprehensive Nuclear Test-Ban Treaty (CTBT), the International Monitoring System (IMS) was established to monitor the world for nuclear weapon explosions. As part of this network, systems are in place to monitor the atmosphere for radioxenon. The IMS routinely detects radioxenon from sources other than nuclear explosions. One of these radioxenon sources is radiopharmaceutical production facilities. This is a sensitivity study on the nuclear forensic signals possible from such facilities. A fission process model was produced to calculate the activity of ^131mXe, ^133mXe, ¹³³Xe and ¹³⁵Xe in the process utilized to produce ⁹⁹Mo and ¹³¹I for medical applications through high enriched uranium fission. The computer model accounts for fractionation of radionuclides within a decay chain that may result from filtering or chemical procedures. Ratios of the radioxenon isotopes are calculated as a function of decay time after the release. The ratios are then compared to those expected from nuclear explosions. The main conclusion from this work is that the two main factors that affect the nuclear forensic signal from radiopharmaceutical production facilities are the sample irradiation time and the use of emission gas storage tanks.     

Practicality of Tetragonal Nano-Zirconia as a Prospective Sorbent in the Preparation of 99Mo/99mTc Generator for Biomedical Applications

The feasibility of using tetragonal nano-zirconia (t-ZrO₂) as an effective sorbent for developing a ⁹⁹Mo/^99mTc chromatographic generator was demonstrated. The structural characteristics of the sorbent matrix were investigated by different analytical techniques such as XRD, BET surface area analysis, FT-IR, TEM etc. The material synthesized was nanocrystalline, in tetragonal phase with an average particle size of ~7 nm and a large surface area of 340 m² g^?1. The equilibrium sorption capacity of t-ZrO₂ is >250 mg Mo g^?1. The present study indicates that ⁹⁹Mo is both strongly and selectively retained by t-ZrO₂ at acidic pH and ^99mTc could be readily eluted from it, using 0.9% NaCl solution. A 9.25 GBq (250 mCi) t-ZrO₂ based chromatographic ⁹⁹Mo/^99mTc generator was developed and its performance was repeatedly evaluated for 10 days. ^99mTc could be eluted with >85% yield having acceptable radionuclidic, radiochemical and chemical purity for clinical applications. The compatibility of the product in the preparation of ^99mTc labeled formulations such as ^99mTc-EC and ^99mTc-DMSA was evaluated and found to be satisfactory.     

Biosorption of 241Am by immobilized Saccharomyces cerevisiae

More than half of the world's annual production of radionuclides is used for medical purposes such as diagnostic imaging of diseases and patient therapy. Using aqueous homogeneous solution reactor technology, production quantities of medical radioisotopes ⁹⁹Mo and⁸⁹Sr, can be extracted from one reactor cycle. ⁹⁹Mo may be produced directly from UO₂SO₄ uranyl sulfate in an aqueous homogeneous solution nuclear reactor in a manner that produces high purity radionuclides, making efficient use of the reactor's uranium fuel solution. The process is relatively simple, economical, and waste free, eliminating uranium targets. The short-lived radioisotope ^99mTc is eluted from ⁹⁹Mo for diagnostic imaging. Radioisotope ⁸⁹Sr infusion is a therapeutic modality that reduces reliance on narcotic analgesia through palliation of metastatic bone pain caused by metastases of the cancer to the bone. Painful disseminated osseous metastases are common with carcinomas of the lung, prostate, and breast. Synergistic interleaving of two manufacturing processes, one producing ⁹⁹Mo and another producing ⁸⁹Sr in the same production cycle of an aqueous homogeneous solution reactor makes full and efficient use of the time for both the neutron irradiation stage and the extraction stage of each radionuclide. Interleaving the capture of ⁸⁹Sr radioisotope with production processing of ⁹⁹Mo radioisotope is achieved, since the extraction and subsequent elimination of radionuclide impurities occurs during separate parts of the reactor cycle. The process applies to either HEU or LEU nuclear fuels in an aqueous homogeneous solution reactor.