Ashutosh Dash

Sustained Availability of 99mTc: Possible Paths Forward

The availability of 99mTc for single-photon imaging in diagnostic nuclear medicine is crucial, and current availability is based on the 99Mo/99mTc generator fabricated from fission-based molybdenum (F 99Mo) produced using high enriched uranium (HEU) targets. Because of risks related to nuclear material proliferation, the use of HEU targets is being phased out and alternative strategies for production of both 99Mo and 99mTc are being evaluated intensely. There are evidently no plans for replacement of the limited number of reactors that have primarily provided most of the 99Mo. The uninterrupted, dependable availability of 99mTc is a crucial issue. For these reasons, new options being pursued include both reactor- and accelerator-based strategies to sustain the continued availability of 99mTc without the use of HEU. In this paper, the scientific and economic issues for transitioning from HEU to non-HEU are also discussed. In addition, the comparative advantages, disadvantages, technical challenges, present status, future prospects, security concerns, economic viability, and regulatory obstacles are reviewed. The international actions in progress toward evolving possible alternative strategies to produce 99Mo or 99mTc are analyzed as well. The breadth of technologies and new strategies under development to provide 99Mo and 99mTc reflects both the broad interest in and the importance of the pivotal role of 99mTc in diagnostic nuclear medicine.

99Mo/99mTc separation: An assessment of technology options

Several strategies for the effective separation of (99m)Tc from (99)Mo have been developed and validated. Due to the success of column chromatographic separation using acidic alumina coupled with high specific activity fission (99)Mo (F (99)Mo) for production of (99)Mo/(99m)Tc generators, however, most technologies until recently have generated little interest. The reduced availability of F (99)Mo and consequently the shortage of (99)Mo/(99m)Tc column generators in the recent past have resurrected interest in the production of (99)Mo as well as (99m)Tc by alternate routes. Most of these alternative production processes require separation techniques capable of providing clinical grade (99m)Tc from low specific activity (99)Mo or irradiated Mo targets. For this reason there has been renewed interest in alternate separation routes. This paper reviews the reported separation technologies which include column chromatography, solvent extraction, sublimation and gel systems that have been traditionally used for the fabrication of (99)Mo/(99m)Tc generator systems. The comparative advantage, disadvantage, and technical challenges toward adapting the emerging requirements are discussed. New developments such as solid-phase column extraction, electrochemical separation, extraction chromatography, supported liquid membrane (SLM) and thermochromatographic techniques are also being evaluated for their potential application in the changed scenario of providing (99m)Tc from alternate routes. Based on the analysis provided in this review, it appears that some proven separation technologies can be quickly resurrected for the separation of clinical grade (99m)Tc from macroscopic levels of reactor or cyclotron irradiated molybdenum targets. Furthermore, emerging technologies can be developed further to respond to the expected changing modes of (99m)Tc production.

Nanoceria-PAN composite-based advanced sorbent material: a major step forward in the field of clinical-grade 68Ge/68Ga generator.

The (68)Ge/(68)Ga generator has high potential for clinical positron emission tomography (PET) imaging. However, because of the unavailability of a suitable sorbent material, the commercially available (68)Ge/(68)Ga generators are not directly adaptable for the preparation of (68)Ga-labeled radiopharmaceuticals. In view of this, a new nanoceria-polyacrylonitrile (PAN) composite sorbent has been synthesized by decomposition of a cerium oxalate precursor to cerium oxide and its subsequent incorporation in PAN matrix for the development of a clinical grade (68)Ge/(68)Ga generator. The X-ray diffraction (XRD) studies and BET nitrogen adsorption technique revealed that nanocrystalline ceria had an average particle size of approximately 10 nm, surface area of 72 +/- 3 m(2)/g and an average pore size of 3.8 +/- 0.1 A. Investigation of the distribution ratio (K(d)) values for the prepared sorbent in 0.01 N HCl medium revealed the suitability of the sorbent for the quantitative retention of (68)Ge and efficient elution of clinical grade (68)Ga. A 370 MBq (10 mCi) (68)Ge/(68)Ga chromatographic generator was developed using this sorbent. (68)Ga could be regularly eluted from this generator with >80% elution yield. The eluted (68)Ga possess high radionuclidic purity (<1 x 10(-5)% of (68)Ge impurity), chemical purity (<0.1 ppm of Ce, Fe and Mn ions) and was amenable for the preparation of (68)Ga-labeled radiopharmaceuticals. The generator gave a consistent performance with respect to the elution yield and purity of (68)Ga over an extended period of 7 months.

Development of a nano-zirconia based 68Ge/68Ga generator for biomedical applications.

INTRODUCTION Most of the commercially available (68)Ge/(68)Ga generator systems are not optimally designed for direct applications in a clinical context. We have developed a nano-zirconia based (68)Ge/(68)Ga generator system for accessing (68)Ga amenable for the preparation of radiopharmaceuticals. METHODS Nano-zirconia was synthesized by the in situ reaction of zirconyl chloride with ammonium hydroxide in alkaline medium. The physical characteristics of the material were studied by various analytical techniques. A 740 MBq (20 mCi) (68)Ge/(68)Ga generator was developed using this sorbent and its performance was evaluated for a period of 1 year. The suitability of (68)Ga for labeling biomolecules was ascertained by labeling DOTA-TATE with (68)Ga. RESULTS The material synthesized was nanocrystalline with average particle size of ~7 nm, pore-size of ~4 Å and a high surface area of 340±10 m(2) g(-1). (68)Ga could be regularly eluted from this generator in 0.01N HCl medium with an overall radiochemical yield >80% and with high radionuclidic (<10(-5)% of (68)Ge impurity) and chemical purity (<0.1 ppm of Zr, Fe and Mn ions). The compatibility of the product for preparation of (68)Ga-labeled DOTA-TATE under the optimized reaction conditions was found to be satisfactory in terms of high labeling yields (>99%). The generator gave a consistent performance with respect to the elution yield and purity of (68)Ga over a period of 1 year. CONCLUSIONS The feasibility of preparing an efficient (68)Ge/(68)Ga generator which can directly be used for biomedical applications has been demonstrated.

Rhenium-188: availability from the (188)W/(188)Re generator and status of current applications.

Rhenium-188 is one of the most readily available generator derived and useful radionuclides for therapy emitting β(-) particles (2.12 MeV, 71.1% and 1.965 MeV, 25.6%) and imageable gammas (155 keV, 15.1%). The (188)W/(188)Re generator is an ideal source for the long term (4-6 months) continuous availability of no carrier added (nca) (188)Re suitable for the preparation of radiopharmaceuticals for radionuclide therapy. The challenges associated with the double neutron capture route of production of the parent (188)W radionuclide have been a major impediment in the progress of application of (188)Re. Tungsten-188 of adequate specific activity can be prepared only in 2-3 of the high flux reactors operating in the World. Several useful technologies have been developed for the preparation of clinical grade (188)W/(188)Re generators. Since the specific activity of (188)W used in the generator is relatively low 185 GBq( < 5 Ci)/g], the eluted (188)ReO(4)(-) can have low radioactive concentration often insufficient for radiopharmaceutical preparation. However, several efficient post elution concentration techniques have been developed that yield clinically useful (188)ReO(4)(-) solutions. Rhenium-188 has been used for the preparation of therapeutic radiopharmaceuticals for the management of diseases such as bone metastasis, rheumatoid arthritis and primary cancers. Several early phase clinical studies using radiopharmaceuticals based on (188)Re-labeled phosphonates, antibodies, peptides, lipiodol and particulates have been reported. This article reviews the availability and use of (188)Re including a discussion of why broader use of (188)Re has not progressed as expected as a popular radionuclide for therapy.

Detailed evaluation on the effect of metal ion impurities on complexation of generator eluted 68Ga with different bifunctional chelators.

INTRODUCTION The introduction of (68)Ga-based positron emission tomography (PET) to clinical practice using (68)Ge/(68)Ga generator represents a developmental milestone in the field of molecular imaging. Herein, we report a systematic study on (68)Ga complexes with different bifunctional chelators (BFCs) and the effect of metal ion impurities on the radiochemical yields in order to identify the most suitable BFC to be used for the development of (68)Ga-based target specific radiopharmaceuticals. METHODS Radiolabeling of four commonly used BFCs namely p-isothiocyanato benzyl derivatives of diethylenetriaminepentacetic acid (DTPA), 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid (DOTA), 1,4,7-triazacyclononane-1,4,7-triacetic acid (NOTA) and 3,6,9,15-tetraazabicyclo [9.3.1]pentadeca-1(15),11,13-triene-3,6,9-triacetic acid (PCTA) with (68)Ga was studied with respect to optimal radiolabeling conditions, effect of metal ion impurities on radiochemical yield, in vitro stability and in vivo clearance properties in biological system. RESULTS Out of the four BFCs studied, p-isothiocyanato benzyl-1,4,7-triazacyclononane-1,4,7-triacetic acid (NOTA-NCS) could be radiolabeled instantly with (68)Ga at room temperature with >98% yield, even in presence of up to 10 ppm of other metal ion impurities (such as Zn, Cu, Fe, Al, Sn and Ti ions). The (68)Ga-complex of NOTA-NCS demonstrated high in vitro stability even in the presence of 1000 times molar excess of metal ions (such as Fe, Cu, Zn and Ca ions). In contrast, other (68)Ga-labeled BFCs (DTPA-NCS, DOTA-NCS and PCTA-NCS) showed reduced radiochemical yields when incubated with the above concentration of metal ions. The biodistribution studies in Swiss mice revealed that (68)Ga-NOTA-NCS cleared rapidly through the kidneys with minimum retention in any major organ. CONCLUSIONS The simple and rapid approach for preparation of (68)Ga-radiopharmaceuticals using NOTA based bifunctional chelators would render (68)Ga-radiopharmaceutical chemistry more convenient with minimum interference from other metal ion impurities; and increase the scope of making (68)Ga based agents for PET imaging.

Availability of yttrium-90 from strontium-90: a nuclear medicine perspective.

Yttrium-90 (T(½) 64.1 hours, E(βmax)=2.28 MeV) is a pure β⁻ particle emitting radionuclide with well-established applications in targeted therapy. There are several advantages of ⁹⁰Y as a therapeutic radionuclide. It has a suitable physical half-life (∼64 hours) and decays to a stable daughter product ⁹⁰Zr by emission of high-energy β⁻ particles. Yttrium has a relatively simple chemistry and its suitability for forming complexes with a variety of chelating agents is well established. The ⁹⁰Sr/⁹⁰Y generator is an ideal source for the long-term continuous availability of no-carrier-added ⁹⁰Y suitable for the preparation of radiopharmaceuticals for radionuclide therapy. The parent radionuclide ⁹⁰Sr, which is a long-lived fission product, is available in large quantities from spent fuel. Several useful technologies have been developed for the preparation of ⁹⁰Sr/⁹⁰Y generators. There are several well-established radiopharmaceuticals based on monoclonal antibodies, peptides, and particulates labeled with ⁹⁰Y, that are in regular use for the treatment of some forms of primary cancers and arthritis. At present, there are no generators for the elution of ⁹⁰Y that can be set up in a hospital radiopharmacy. The radionuclide is procured from manufacturers and the radiopharmaceuticals are formulated on site. This article reviews the development of ⁹⁰Sr/⁹⁰Y generator and the development of ⁹⁰Y radiopharmaceuticals.

Targeted Radionuclide Therapy - An Overview

Radionuclide therapy (RNT) based on the concept of delivering cytotoxic levels of radiation to disease sites is one of the rapidly growing fields of nuclear medicine. Unlike conventional external beam therapy, RNT targets diseases at the cellular level rather than on a gross anatomical level. This concept is a blend of a tracer moiety that mediates a site specific accumulation followed by induction of cytotoxicity with the short-range biological effectiveness of particulate radiations. Knowledge of the biochemical reactions taking place at cellular levels has stimulated the development of sophisticated molecular carriers, catalyzing a shift towards using more specific targeting radiolabelled agents. There is also improved understanding of factors of importance for choice of appropriate radionuclides based on availability, the types of emissions, linear energy transfer (LET), and physical half-life. This article discusses the applications of radionuclide therapy for treatment of cancer as well as other diseases. The primary objective of this review is to provide an overview on the role of radionuclide therapy in the treatment of different diseases such as polycythaemia, thyroid malignancies, metastatic bone pain, radiation synovectomy, hepatocellular carcinoma (HCC), neuroendocrine tumors (NETs), non-Hodgkins lymphoma (NHL) and others. In addition, recent developments on the systematic approach in designing treatment regimens as well as recent progress, challenges and future perspectives are discussed. An examination of the progress of radionuclide therapy indicates that although a rapid stride has been made for treating hematological tumors, the development for treating solid tumors has, so far, been limited. However, the emergence of novel tumor-specific targeting agents coupled with successful characterization of new target structures would be expected to pave the way for future treatment for such tumors.

An electro-amalgamation approach to isolate no-carrier-added 177Lu from neutron irradiated Yb for biomedical applications

INTRODUCTION A novel two-step separation process for the production of no-carrier-added (NCA) (177)Lu from neutron irradiated Yb target through an electrochemical pathway employing mercury-pool cathode has been developed. METHODS A two-cycle electrolysis procedure was adopted for separation of (177)Lu from (177)Lu/Yb mixture in lithium citrate medium. The influence of different experimental parameters on the separation process was investigated and optimized for the quantitative deposition of Yb in presence of (177)Lu. The first electrolysis was performed for 50 min in the (177)Lu/Yb feed solution at pH 6 applying a potential of 8 V using platinum electrode as anode and mercury as the cathode. The second electrolysis was performed under the same conditions using fresh electrodes. The radionuclidic and chemical purity of (177)Lu was determined by using gamma ray spectrometry and atomic absorption spectrometry. The suitability of (177)Lu for biomedical applications was ascertained by labeling 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid D-Phe(1)-Tyr(3)-octreotate(DOTA-TATE) with (177)Lu. RESULTS This process could provide NCA (177)Lu with >99.99% radionuclidic purity and an overall separation yield of ∼99% was achieved within 3-4 h. The Hg content in the product was determined to be <1 ppm. Radiolabeling yield of >98% was obtained with DOTA-TATE under the optimized reaction conditions. CONCLUSIONS An efficient strategy for the separation of NCA (177)Lu, suitable for biomedical applications, has been developed.

Nanocrystalline zirconia: A novel sorbent for the preparation of 188W/188Re generator

Nanocrystalline zirconia, a novel high capacity sorbent material was synthesized and tested for its utility in the preparation of (188)W/(188)Re generators. The structural investigation of the material was carried out using X-ray diffraction, surface area determination, FTIR and TEM micrograph analysis. Various experimental parameters were optimized to separate (188)Re from (188)W. The capacity of the material was found to be approximately 325mgW/g at the optimum pH. A chromatographic (188)W/(188)Re generator was developed using this material from which >80% of (188)Re generated could be eluted with 0.9% saline solution, with high radionuclidic, radiochemical and chemical purity and appreciably high radioactive concentration suitable for radiopharmaceutical applications.