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By Dr. Jeff R. Livingstone, Unisys
Many healthcare organizations already have started to shift from the traditional fee-for-service model to a value-based care (VBC) model in which providers are incentivized to provide proactive, coordinated care in an effort to deliver better patient outcomes.
However, advanced healthcare analytics along with artificial intelligence (AI) represent critical pieces to the future of VBC and will help shift the fundamental landscape of medicine in the coming 12 to 24 months.
At its core, the VBC model represents a shift in mindset – from thinking of the patient as just a person needing treatment to thinking of the patient as a customer buying a product. That means creating an overall positive customer experience is more important than ever. Being able to leverage innovative technology to strengthen a provider’s VBC model is key not only to providing the most effective patient care possible, but to delivering an optimal customer experience as well. And in this aspect, disruptive technology can play a big role.Analytics and AI have great potential to make significant and positive impacts within care management, patient diagnoses and medical intervention.
The striking diffusion and exciting perspectives of 177Lu in targeted radionuclide therapy (TRT) are primarily attributed to the following.
A wide range of 177Lu radiopharmaceuticals has been successfully developed and evaluated. The in vivo applications of key 177Lu radiopharmaceuticals for a variety of therapeutic procedures include peptide receptor radionuclide therapy [11–26], bone pain palliation [27–33], radiation synovectomy [34–39] and radioimmonutherapy [40–46]. There is a steadily expanding list of 177Lu-labeled radiopharmaceuticals that is currently being evaluated at the preclinical research or at product development stages; these may potentially be used in vivo in humans for evaluation for radionuclide therapy [1–3]. A summary of key 177Lu-labeled radiopharmaceuticals currently used in TRT is depicted in Table 1.
The information comes fromhttps://www.ncbi.nlm.nih.gov/pmc/articles/PMC4463871
Technetium-99m is a radioactive tracer that is used in twenty million medical diagnostic procedures per year. At least 31 radiopharmaceuticals based on Tc-99m are used for imaging and studying organs such as the brain, heart muscle, thyroid, lungs, liver, gallbladder and kidneys, as well as the skeleton and blood and for the investigation of tumours.
The ‘m’ in the name of technetium-99m indicates that it is metastable. Tc-99m is radioactive because one or more of the protons and neutrons in its nucleus is in an excited state. Tc-99m decays into Tc-99 with a half-life of six hours and this makes it particularly well suited to use in the body: after one day (four half-lives) only 6.3% of the initial Tc-99m remains. (It’s worth noting that the non-metastable technetium-99 is also radioactive, but with a half-life of 211000 years, it presents a very low risk.)
This short half-life also creates a problem: obtaining Tc-99m when required. Hospitals cannot run their own nuclear reactors and so they rely on technetium generators – machines that produce Tc-99m from the decay of its parent isotope molybdenum-99. Molybdenum-99 has a longer half-life (66 hours) and can therefore be transported to hospitals and still remain useful for up to a week.
Molybdenum-99 is produced in nuclear reactors by bombarding a highly enriched uranium target with neutrons, causing it to fission, forming Mo-99 (and many other isotopes) as it does. The vast majority of Mo-99 is produced by five nuclear reactors around the world that are specifically devoted to the production of nuclear isotopes for medicine: NRU in Canada, BR2 in Belgium, SAFARI-1 in South Africa, HFR Petten in the Netherlands and OSIRIS-1 in France.* Temporary shutdowns of NRU and HFR Petten in the 2000s led to a long-term shortage of Mo-99.
Once Mo-99 has been produced it is placed into a technetium generator and these generators are transported to hospitals. The technetium generators make use of the fact that molybdenum likes to bond with aluminium oxide (alumina) but technetium does not. The generators are “milked” by drawing a saline solution across an inner molybdenum/alumina capsule; during this elution process any technetium that has formed will be drawn away with the saline and can then be used in tests.A cutaway model of a technetium generator.The molybdenum/alumina sample is placed in the centre of the device, surrounded by shielding (painted red in this case). Saline is injected through one of the tubes at the top of the device and flows into a shielded container through the other tube, after having passed over the sample and “picked up” radioactive technetium-99m.
Alpha particle-emitting isotopes are being investigated in radioimmunotherapeutic applications because of their unparalleled cytotoxicity when targeted to cancer and their relative lack of toxicity towards untargeted normal tissue. Actinium-225 has been developed into potent targeting drug constructs and is in clinical use against acute myelogenous leukemia. The key properties of the alpha particles generated by 225Ac are the following: i) limited range in tissue of a few cell diameters; ii) high linear energy transfer leading to dense radiation damage along each alpha track; iii) a 10 day half-life; and iv) four net alpha particles emitted per decay. Targeting 225Ac-drug constructs have potential in the treatment of cancer.
The treatment of cancer with targeted radionuclide therapy is a maturing field that has achieved significant success with the introduction of two FDA approved drugs adding new modalities to the current cancer therapy arsenal of surgery, chemotherapy and external beam radiation. The specificity imparted by the targeting vehicle (e.g., antibodies (IgGs), peptides, small molecules) can be complemented by attaching a cytotoxic payload to yield a potent therapeutic effect. Particle-emitting radionuclides are some of the most promising cytotoxic moieties when linked to tumor-targeting carrier molecules. To date, much of the work has been done with beta particle-emitting isotopes. Iodine-131 Tositumomab (Bexxar) and Yttrium–90 Ibritumomab Tiuxetan (90Y-Zevalin) are FDA approved beta particle-emitting IgGs used to treat B-cell non-Hodgkin’s lymphoma .
The increased availability and improved radiochemistry of alpha particle-emitting nuclides for targeted therapy have presented novel possibilities for their use in radioimmunotherapy (RIT). Alpha particles offer key advantages over beta particles, in particular are the high linear energy transfer (LET) and the limited range in tissue. The high alpha particle LET is on the order of 100 keV/μm and it can produce substantially more lethal double strand DNA breaks per alpha track than beta particles when traversing a cell nucleus. The alpha particle tracks are relatively short and thus have a limited range in tissue (on the order of a few cell diameters). This confines the toxic effect to a relatively small field - within a few cell diameters from the site of decay versus the much longer-ranged beta particles. 90Y for example, has a maximum range on the order of several hundred cell diameters and thus deposits energy in the tumor as well as the surrounding normal tissue. The number of particle track transversals through a tumor cell nucleus that was necessary to kill the cell was considerably lower for alpha particles than for beta particles and it has been estimated that one alpha particle transversal can kill a cell . This higher biological effectiveness seems nearly independent of oxygen concentration, dose rate and cell cycle position.
Preclinical research has demonstrated the potential of alpha particle-emitting isotopes in RIT [3,4]. Alpha emitting nuclides displayed cytotoxicity in a model of leukaemia that was resistant to beta- and gamma-radiotherapy and doxorubicin chemotherapy . There are a number of alpha particle-emitting nuclides considered for application in targeted therapy displaying half-lives ranging from minutes to days. 213Bi was one alpha particle-emitting nuclide (t1/2 = 46 min) that has been proposed for therapeutic use and has been evaluated clinically. However, 213Bi is generator-produced and has a very short half-life. The clinical use of 213Bi presents the logistical dilemma of eluting the generator, radiolabeling the targeting molecule, administering a dose, and allowing sufficient time for targeting. All of these steps consume valuable time that decreases the effective dose administered. An innovative alternative to 213Bi was the use of its parent nuclide, 225Ac, which has a ten-day half-life and 4 net alpha particle-emissions per decay. In vitro cytotoxicity data using the same antibodies labeled with either 213Bi or 225Ac demonstrated that several logs less 225Ac radioactivity was necessary to reach LD50, presumably because of the multiple alpha emissions and the 300-fold longer half-life . The enhanced potency of 225Ac versus 213Bi was also directly demonstrated in a murine model of human prostate cancer [6,7]. This article reviews the literature of 225Ac dealing with its production and supply, physical, chemical and biological properties, dosimetry, and clinical use as a radiotherapeutic agent in cancer therapy.
Actinium was aptly named for the Greek aktis or aktinos, meaning ray or beam . The element was discovered by Debierne in 1899 and Giesel in 1902 . Actinium occurs naturally in association with uranium radionuclides and 225Ac can be obtained either from the decay of 233U or from the neutron transmutation of 226Ra by successive n,γ capture decay reactions via 227Ac, 228Th to 229Th [10–12]. Currently, there are two sources of 225Ac that have been used in clinical trials: (1) the U.S. Department of Energy, Oak Ridge National Laboratory (ORNL) in Oak Ridge, TN, United States of America and (2) the Institute for Transuranium Elements in Karlsruhe, Germany. The 225Ac at both sites was derived from 233U that was produced as a component of the U. S. molten salt breeder reactor program [13–15], and had been in long-term storage at ORNL. The bulk of this high purity and low specific activity 229Th was separated from waste material associated with the original production of the 233U. This 229Th yields 225Ac that was produced as a “carrier-free” nuclide and was suitable for use in clinical research applications. The 225Ac from both sources has been used to construct 213Bi producing generators [16,17] for Phase I and I/II clinical treatment of leukemia  and the 225Ac from Oak Ridge used to directly radiolabel an antibody for application in a Phase I clinical trial treating leukemia (ongoing clinical trial ).
An 225Ac generator based on a design that adsorbed 229Th oxide onto a titanium phosphate resin was described by Geerlings et al. . Elution of this 229Th cow with dilute nitric acid yielded a mixture of radionuclides: 225Ac, 225Ra, and 224Ra. Another downstream column, containing Dowex 50 WX8, was used to purify the 225Ac by removing the 225Ra, 224Ra, and the 224Ra decay products. In 1993, it was proposed that the 225Ac thus produced could be used to label an antibody or be affixed to a resin as a parent for a 213Bi generator product.
The separation method used at Oak Ridge National Laboratory to isolate 225Ac from the 229Th stock allows 229Th, 225Ra, and 225Ac to reach equilibrium (45 d) and then carrier-free 225Ra and 225Ac were separated from the thorium stock in nitric acid using anion exchange chromatography . Concentration of the 225Ra/225Ac eluate was effected by evaporation or neutralization and co-precipitation. Recently, this group has described a four-step chemical separation procedure, employing both anion and cation exchange chromatography, to process their current supply of 150 mCi of 229Th into 225Ac . Over an 8-week period, approximately 100 mCi of 225Ac was yielded per processing campaign and the product shipped in 5–6 batches. Following the initial process run of a campaign yielding ~ 50–60 mCi 225Ac, the radium pool was reprocessed bi-weekly to yield the additional 225Ac shipments. This material is currently being utilized in both a 213Bi- and an 225Ac-labeled antibody trial at Memorial Sloan-Kettering Cancer Center (MSKCC) . The average radionuclidic purity was 99.6% ± 0.7% 225Ac with ≤ 0.6% 225Ra contaminant and an average 229Th content of 4 (+5/−4) × 10−5 %.
The separation and purification method to yield 225Ac from a 229Th source that is currently employed at the Institute for Transuranium Elements entails a 229Th stock is batch loaded in nitric acid onto 0.5 L of Dowex anion exchange resin . This process utilizes a combination of extraction and ion exchange chromatographic methods to obtain carrier-free, clinical quality 225Ac with > 95% overall yield. Based upon their stock of 215 mg of 229Th, they can isolate 43 mCi of 225Ra and 39 mCi of 225Ac every 9 weeks. This 225Ac has been used in clinical generators  to produce 213Bi for radiolabeled radioimmunotherapeutic (RIT) pharmaceuticals at MSKCC  and in collaborations with a number of other sites .
A liquid 229Th/225Ac generator used a process that entails maintaining a stock 229Th solution in an ammonium citrate solution in order to eliminate the radiolysis and degradation experienced with solid sorbents . As the 225Ra and 225Ac reach equilibrium with the 229Th, they are isolated in a one-step cation exchange process. The 229Th breakthrough was effectively removed in a single separation cycle by changing the pH of the solution. This process takes advantage of the differences in the stability constants of thorium (K1 = 1013, K2 = 108) and actinium (K1 ~ K2 = 106) citrate complexes .
Cyclotron production is an alternate strategy for 225Ac production that employs proton irradiation of 226Ra leading to 225Ac via [p,2n] reactions [10,11,23]. Theoretically, irradiation of 1 mg of 226Ra should yield 35 mCi of 225Ac . Recently, the feasibility of cyclotron produced 225Ac was demonstrated and maximum yields were reached with an incident proton energy of 16.8 MeV  using the 226Ra(p,2n)225Ac reaction. In this work, 0.0125 mg of 226Ra yielded 0.0021 mCi 225Ac after irradiation of a 36 mm2 target with a 10 μA proton current for 7 h. No significant differences were found in the radionuclidic purity of the cyclotron product when compared to 225Ac produced via the 229Th method  and 213Bi produced from this 225Ac was found to label antibody constructs with approximately 90% yield.
225Ac decay (see Figure 1) yields six principal radionuclide progeny in the decay cascade to stable 209Bi . A single 225Ac (t1/2 = 10.0 d; 6 MeV α particle) decay yields net 4 alpha and 3 beta disintegrations, most of high energy and 2 useful gamma emissions of which the 213Bi 440 keV γ emission has been used in imaging drug distribution . These daughters are 221Fr (t1/2 = 4.8 m; 6 MeV α particle and 218 keV γ emission), 217At (t1/2 = 32.3 ms; 7 MeV α particle), 213Bi (t1/2 = 45.6 m; 6 MeV α particle, 444 keV β− particle and 440 keV γ emission), 213Po (t1/2 = 4.2 μs; 8 MeV α particle), 209Tl (t1/2 = 2.2 m; 659 keV β− particle), 209Pb (t1/2 = 3.25 h; 198 keV β− particle) and 209Bi (stable). Given the 10.0 d half-life of 225Ac, the large alpha particle emission energies, and the favorable rapid decay chain to stable 209Bi this radionuclide was recognized as a potential candidate for use in cancer therapy . Figure 1 illustrates the decay scheme of 225Ac.
The 225Ac decay scheme.
The potential for using 225Ac as a therapeutic radionuclide was limited many years by the paucity of suitable chelating moieties capable of stably binding this radionuclide as well as controlling the fate of the daughters . Additionally, the chemistry of actinium was not explored or well developed. Diamond and Seaborg studied the elution profiles of the transuranium elements in hydrochloric acid on cation-exchange resin and concluded that the actinides may form complex ions with chloride to a greater extent than the lanthanide elements based on the partial covalent character of the actinide bonds involved in the hybridization of the 5f orbitals . The actinium(III) ionic radius was reported as 0.111 nm . Radiopolarographic reduction of the 225Ac(III) ion in aqueous solution in the presence of 1,4,7,13,16-hexaoxacyclooctadecane (18-CROWN-6) and suggested the formation of a divalent actinium cation . In the absence of 18-CROWN-6, the measured E1/2 value was −2.15 V versus SCE and as increasing concentrations of 18-CROWN-6 were added, the E1/2 value shifted to a more negative potentials in a linear fashion. They concluded from this study that the Ac(II) ionic radius was 0.125 nm and the electronic configuration was [Rn]6d1. The overall hydrolytic constant, β3, for the hydrolysis of 225Ac(III) in aqueous NaClO4 (μ = 0.1) solution was determined using the electromigration method . A plot of the 225Ac ion velocity as a function of pH showed a constant velocity of 5.4 × 10−4 cm2/Vs in the pH interval from 4–10, indicating that no hydrolytic process occurred until pH = 10. At pH 10, the velocity dropped steeply and by pH 11 the velocity was 0. A value of pβ3 = 31.9 ± 0.2 was calculated for the hydrolytic reaction, Ac3+ + 3H2O → Ac(OH)3 + 3H+, where β3 was the hydrolytic constant. When the pH was less than 4, a 10–15% decrease in ion mobility was measured.
Attempts to utilize 225Ac as a tumoricidal agent steered the evaluation of a different complexing agents and chelates in order to enhance tumor uptake and avoid normal organ uptake. 225Ac complex stability improved by trial and error and a trend was recognized as one moved from simple complexing agents to acyclic chelates to macrocyclic chelates. Two related macrocyclic chelates, in particular, were identified as potentially useful and further explored as moieties to attach to targeting monoclonal antibody carriers. The first was 1,4,7,10,13,16-hexaazacyclohexadecane-N,N′,N″,N‴,N′‴,N″‴-hexaacetic acid (HEHA) and the second was 1,4,7,10-tetraazacyclododecane-N,N′,N″,N‴-tetraacetic acid (DOTA). These chelates are related because they are both macrocycles that present carboxylic acid and amine functionalities to the metal-ion, albeit with different denticity, macrocycle size, and overall charge. As will be described below, the 225Ac complex with HEHA demonstrated less stability than the 225Ac complex with DOTA in vivo in the experiments described. In addition, the two monoclonal antibody/antigen systems that were examined using HEHA-constructs were non-internalizing immune complexes and the targeted constructs could still release the daughters systemically. The 225Ac released from the HEHA was distributed to liver and bone and the subsequent release of its daughters contributed to acute radiotoxicity as did the daughters from the targeted, but not internalized parent. The 225Ac complex formed with DOTA was considerably more stable in vivo and the antibodies selected for these studies formed internalizing immune complexes with their respective antigen targets.
Isothiocyanate-functionalized-DOTA derivatives were selected as the most promising to pursue for coupling to antibody molecules from out of a group of potential 225Ac chelate compounds: diethylenetriaminepentaacetic acid (DTPA), 1,4,8,11-tetraazacyclotetradecane-1,4,8,11-tetraacetic acid (TETA), DOTA, 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetrapropionic acid (DOTPA), 1,4,8,11-tetraazacyclotetradecane-1,4,8,11-tetrapropionic acid (TETPA), and 1,4,7,10-tetraazacyclododecane-1,4,7,10-tetramethylenephosphonic acid (DOTMP). The bifunctional chelating agents MeO-DOTA-NCS, (α-(5-isothiocyanato-2-methoxyphenyl)-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid and 2B-DOTA-NCS, 2-(p-isothiocyanatobenzyl)-1,4,7,10-tetraazacyclododecane-1,4,7,10-tetraacetic acid were both evaluated and compared as their respective [225Ac]DOTA-antibody construct. A two-step labeling method was developed using several different IgG systems . The chelation reaction yield in the first step was 93% ± 8% radiochemically pure. The second step, which couples the [225Ac]DOTA-SCN moiety to the IgG, yielded constructs that were 95% ± 5% radiochemically pure and with mean percent immunoreactivity ranging from 25–81%, depending on the antibody used but consistent for each IgG system. This methodology met the requirement for high temperature labeling of the DOTA chelate with 225Ac without sacrificing the biological activity of the protein.
The biodistribution of a mixed 225Ac, 169Yb and 148Pm sodium citrate solution (pH 6.5) was examined in normal rats and mice bearing adenocarcinoma tumor implants . The rats were injected i.v. while the mice were injected intraperitoneally (i.p.). Each animal received 40 kBq (1080 nCi) of 225Ac in a citrate solution. Animals were euthanized 5 h post-injection and blood, liver, femur, urine, and tumor were harvested. Samples were measured 1 d after harvest after secular equilibrium was established. As expected for a rapidly clearing small molecular weight metal-citrate complex, the percent injected dose (%ID) of 225Ac per gram of blood was low, 0.06 and 1.2 for rats and mice, respectively. Inter-species liver and femur values were different; the rats had 5.7 and 1.2 %ID 225Ac/g while the mice had 38.7 and 16.8 %ID 225Ac/g, respectively. Tumor tissue in mice accumulated 3.5 %ID 225Ac/g.
The influence of varying ethylenediaminetetramethylenephosphonic acid (EDTMP) solution concentrations on the biodistribution of 225Ac was determined in tumor-bearing mice . Swiss nude mice were implanted with subcutaneous (s.c.) T380 human colon carcinoma and injected i.v. via the tail vein with approximately 50 kBq 225Ac (1350 nCi) in formulations varying the concentration (0, 0.01, 0.05, 0.1, 0.2, 0.5, 1, 2, 10 and 30 mM) of EDTMP or in a 1 mM citrate control solution, adjusted to pH 6.5. Animals were sacrificed at 15 h and tissues were harvested and measured using a high resolution Ge-spectrometer after secular equilibrium was established. [225Ac]citrate control and [225Ac]EDTMP solutions (up to 0.1 mM EDTMP) demonstrated high liver uptake (40 %ID/g). The higher EDTMP concentrations exhibited less than 1 %ID 225Ac/g accumulated in the liver, suggesting that the excess EDTMP assisted the clearance of the 225Ac.
The biodistribution, dosimetry and radiotoxicity of 225Ac complexed with acetate, ethylenediaminetetraacetic acid (EDTA), 1,4,7,10,13-pentaazacyclopentanadecane-N,N′,N″,N‴, N′‴-pentaacetic acid (PEPA), and the A″ isomer of N-[(R)-2-amino-3-(4-nitrophenyl)propyl]-trans-(S,S-cyclohexane-1,2-diamine-N,N,N′,N″,N″-pentaacetic acid (CHX-A″-DTPA) was examined in female BALB/c mice . Animals were sacrificed at 1h, 4h, 24h, 5d, and 8d following i.v. tail vein injection of 92 kBq (2500 nCi) of each of the 225Ac complexes. Tissue samples were harvested, held for 4 h, and then counted using a NaI(Tl) γ-scintillation counter. Data expressed as the %ID/g again demonstrated that the liver was the major site of 225Ac localization for all four small molecule complexes studied. Liver accumulation increases according to the decreasing strength of the 225Ac-complex: CHX-DTPA ~ PEPA > EDTA > acetate. For example, the 24h liver biodistribution data values were approximately 13, 15, 53, and 110 %ID/g for the respective [225Ac]CHX-DTPA, [225Ac]PEPA, [225Ac]EDTA, 225Ac-acetate complexes. When the data were expressed as the % injected dose per organ it was shown that for [225Ac]CHX-DTPA, the bone was the predominant localization site and the liver next (at 24 h, 42 %ID/bone and 13 %ID/liver). Liver and femur accumulation presumably resulted from the loss of 225Ac from the chelators. Absorbed dose values for 225Ac were estimated based upon the data from [225Ac]CHX-DTPA and [225Ac]EDTA. Values are reported as Gy per 92 kBq of injected dose of 225Ac-complex. For [225Ac]CHX-DTPA and [225Ac]EDTA the doses to liver were 30.4 and 117.8; 7.8 and 15.4 to bone; and 3.2 and 2.1 to kidney, respectively. Both acute and chronic toxicity were assessed by organ system damage and white blood cell (WBC) counts as a function of the dose administered. Two mice receiving 92, 185, 370, 740 kBq (2,500, 5,000, 10,000, 20,000 nCi) of [225Ac]CHX-DTPA were sacrificed at 2, 5, 7, 53 d post-injection (the latter two time-points were included for animals that had not succumbed to radiation toxicity during the study). The control was i.v. injected CHX-DTPA chelate-alone in MES buffer. Tissues from these animals were harvested, H&E stained and fixed, evaluated histopathologically, and graded for radiation damage. The tissues that showed the most radiation-induced toxicity were the bone marrow, spleen, gastrointestinal tract, and the liver. At the 92 kBq dose level, the WBC, spleen and bone marrow were rated as having loss of cellular numbers, integrity, orientation, or structure. At the 185 kBq dose level the WBC, spleen, bone marrow, liver, GI tract, and kidney all were rated as having loss of cellular numbers, integrity, orientation, or structure and evidence of cellular necrosis .
Another series of 225Ac-labeled chelate complexes was prepared and their biodistribution measured in normal BALB/c mice . One of these complexes was reported to exhibit improved in vivo stability relative to the others in the series examined. The chelates studied were EDTA, CHX-A-DTPA, PEPA, DOTA, HEHA, and acetate. 92.5 kBq of each complex (2500 nCi) in MES buffer at pH 6.2 was injected into normal female BALB/c mice via the tail vein. Biodistributions were performed at 1, 4, 24, 120 h and samples were counted after 4 h to allow for secular equilibrium. All of the complexes rapidly cleared the blood with < 2 %ID/g in 1 h. The order of most 225Ac distributed into tissue to the least was acetate > EDTA > CHX-A″-DTPA ~ PEPA > DOTA > HEHA. Consistent with the studies described above, the loss of 225Ac from an acyclic chelate was greatest and reflected in the high uptake in liver and bone and poor whole body clearance. [225Ac]HEHA was rapidly excreted within 1 h and only 0.17 % ID/g remained (approximately 100 nCi of the 2500 nCi injected).
The development of tumor specific agents was addressed by preparing antibody constructs with the various chelate candidate molecules identified in the previously described reports. The cytotoxicity of an [225Ac]DTPA-antibody construct was reported in vitro using a murine IgG1 that targets a carbohydrate structure associated with the EGF receptor expressed on the human epidermoid A431 tumor cell line . The specific antibody construct was more potent than the non-specific control construct. However, the DTPA chelate moiety clearly did not stably bind the 225Ac in these experiments as demonstrated by the 225Ac constructs being capable of killing target cells only slightly better than similarly labeled 213Bi-antibodies. In another radiolabeled antibody-DTPA construct study, the biodistribution of [225Ac]DTPA-201B, which targeted murine lung endothelial thrombomodulin, was examined . The construct efficiently targeted the lung but the 225Ac had a very short tissue t1/2 of 4–5 h as compared with the [125I]-201B construct with a t1/2 of 4–5 d. Obviously, the DTPA was unable to stably bind the 225Ac in vivo and this chelate was not going to advance the use of 225Ac in RIT applications.
HEHA is a multidentate, macrocyclic chelate having 6 carboxylic acids and 6 amino nitrogens (compare with PEPA having 5 of each and DOTA having 4 of each). Given the larger macrocycle size, greater number of coordinating ligands, and overall negative charge, the [225Ac]HEHA-antibody complex was hypothesized to be a potentially useful RIT construct. The properties of HEHA were useful for rapidly clearing the simple complex in vivo, but did not address the potential stability should the HEHA chelate be used in an IgG construct that would presumably have a longer biological half-life . The interest in the HEHA chelate led to the description of the synthesis of an isothiocyanate bifunctional chelate (BFC) derivative and the subsequent conjugation to three different antibodies, and radiolabeling of with 225Ac . One of the [225Ac]HEHA-IgG constructs was evaluated for serum stability in fetal bovine serum at 37°C over a 3 d period. After 0.4 h of incubation, 99% of the 225Ac was still associated with the construct, but at 1, 3, 5, 24, and 48 h this value decreased rapidly to 77, 73, 69, < 50, and < 50%, respectively. It was determined that the HEHA rapidly released the 225Ac associated with the targeting antibody and bound to serum proteins.
The [225Ac]HEHA-201B antibody construct was evaluated for vascular targeted therapy of lung tumors in the first reported 225Ac RIT study of in vivo . This study performed biodistribution, dosimetry and therapeutic efficacy studies in female BALB/c mice with the EMT-6 mammary carcinoma as the model. At 1 and 4 h post-injection, 300 %ID/g of 225Ac was distributed to the lung tissue, however, the 225Ac cleared the lung with a t1/2 of 49 h and the released 225Ac accumulated predominantly in the liver, spleen, and bone. It was calculated that a dose of 6 Gy per μCi was delivered to the lungs and about three-fold less to other tissues. A RIT study was performed where 18.5 kBq of construct administered per mouse and 10% of the tumor bearing animals survived 23 d vs. 11 to 15 d for untreated controls. 80% of the animals treated with 37 kBq of 225Ac drug had the tumor eradicated, but died at 16 d from acute radiotoxic effects (total bone marrow ablation, splenic atrophy, damage to the lining of the stomach and intestine). In subsequent studies, no therapeutic window could be identified that would effectively treat tumor but spare the host. The HEHA chelate released the 225Ac and the non-internalizing antibody-antigen complex compromised the use of this system for RIT.
It was hypothesized that if an antibody armed with 225Ac did not form an internalizing antibody-antigen complex, then a smaller domain-deleted fragment of that antibody could better extravasate and penetrate the tumor where the progeny would remain localized as compared to the native, full-sized IgG . The CC49 antibody and the humanized domain-deleted product, ΔCH2CC49, were converted to HEHA-appended constructs, radiolabeled with 225Ac, and evaluated for biodistribution, microdistribution, and therapeutic efficacy in mouse models with s.c. and/or intramuscular (i.m.) LS174T xenografts. The biodistribution data revealed that the [225Ac]HEHA-CC49, [225Ac]HEHA-ΔCH2CC49, and [225Ac]HEHA-control antibody constructs accumulated after 24 h with 25, 18, and 10 %ID/g in the s.c. tumors and 8, 9, and 5 %ID/g of in the i.m. tumors, respectively. Liver and spleen accumulated 225Ac, which increased over a week, presumably due to release from the HEHA chelate. The retention of the progeny was investigated by calculating the ratio of 213Bi to 225Ac in the tumors over an 8 d period and it was found that there was little difference between the CC49 and the domain-deleted fragment. In one RIT study, 800 nCi of [225Ac]HEHA-CC49, [225Ac]HEHA-ΔCH2CC49, or [225Ac]HEHA-control were administered to SCID/LS174T models having both s.c. and i.m. xenografts 9 d post-tumor implant. The maximum tolerated dose (MTD) of the constructs was 800 nCi in these animals and all mice exhibited radiotoxicity by day 6 and were sacrificed on day 8. There was no statistical difference in the tumor sizes in this study based upon treatment. A second RIT study was conducted in a Swiss nude/LS174T model 9 d post tumor implant. Animals had either s.c. or i.m. implants, but not both. Animals with i.m. tumors responded best to treatment with 500 nCi [225Ac]HEHA-CC49, with statistically smaller tumors than those treated with [225Ac]HEHA-ΔCH2CC49, control [225Ac]HEHA-IgG, or cold, unlabeled CC49. Animals with s.c. tumors all responded to treatment with 500 nCi [225Ac]HEHA-CC49, [225Ac]HEHA-ΔCH2CC49, or control [225Ac]HEHA-IgG, but the latter two groups suffered from radiotoxicity. A third RIT study focused on the Swiss nude/LS174T model treated 6 d post tumor implant with 0, 250, or 500 nCi of [225Ac]HEHA-ΔCH2CC49 or 500 nCi of the [225Ac]HEHA-IgG control. Animals had either s.c. or i.m. implant, but not both. There were no statistical differences in the tumor sizes per group in the s.c. implant groups. It was concluded that only marginal therapeutic effect could be attained with the [225Ac]HEHA-CC49 and no therapeutic effect was observed using the domain-deleted fragment, contrary to the penetration/retention hypothesis. RIT with HEHA constructs was limited by radiotoxicity to normal organs.
Controlling the fate of the parent radionuclide and the ensuing progeny would be the key to harnessing the therapeutic potential of 225Ac. The initial step was to identify a suitable chelating agent that would yield stable 225Ac complexes in vivo and thereby shape the pharmacokinetic profile of the parent nuclide by keeping it associated with the targeting carrier immunoglobulin for a long time. Managing the distribution, metabolism and clearance of the daughters was a more daunting task. As described above, the attempts which focused on developing a single chelate moiety to accommodate the parent and the progeny, proved too difficult given the range of different periodic properties of these daughters and using antibody fragments did not enhance tumor penetration and or retention. Previous workers had concluded that 225Ac-antibody constructs were too unstable and that the progeny presented an untenable pharmacological problem. The nanogenerator system altered this paradigm and clearly demonstrated the ability to safely and efficaciously use 225Ac as an extraordinarily potent tumor-selective molecular sized generator in both established solid carcinomas or disseminated cancers.
The approach which was taken that focused on first, stably chelating the 225Ac for delivery in vivo to a target cell; second, internalizing the 225Ac-antibody construct into the target cell; third, retaining the progeny inside the target cell and harnessing their cytotoxic potentials; and fourth, minimizing the loss of the daughters to non-target tissues and thus minimizing systemic radiotoxicity. This strategy was called the 225Ac atomic nanogenerator. DOTA was found to be a stable chelate for 225Ac and was attached to the antibody using a robust thiourea chemical linkage. The targeting agents were internalizing IgGs which stably transported the 225Ac to the cell which bound and modulated the construct into the cell where the progeny were retained and decayed. This approach proved extremely cytotoxic to the targeted cancer cells and largely eliminated systemic toxicity to the host. The 225Ac delivered to the cancer cell was effectively a therapeutic nanogenerator of multiple alpha particle emissions within the target cell .
The first practical application of 225Ac in targeted RIT without any accompanying systemic radiotoxicity utilized the nanogenerator approach . This process was dependent on the stable chelation of the parent 225Ac radionuclide and the efficient delivery and internalization of the construct at the tumor target site. Controlling the [225Ac]DOTA-antibody pharmacokinetics was the key to the use of 225Ac as a therapeutic agent. Further, the progeny retained at the target site harnessed their cytotoxic potentials and contributed to safe and effective tumor therapy. It was discovered that a very stable 225Ac complex with DOTA could be rapidly formed at 60°C. The ensuing [225Ac]DOTA complex could then be coupled to an IgG using the thiourea linkage .
The stability in vitro of [225Ac]DOTA-HuM195 was compared to an analogous 177Lu-labeled construct in 100% human serum, 100% mouse serum, and 25% human serum albumin at 37°C for 15 d. The 225Ac construct displayed stability similar to the 177Lu analogue, with less than 5% loss of 225Ac from the chelate over 15 d. The stability results in all three conditions were similar. Stability in vivo was determined using 10 female nude mice injected i.v. with 300 nCi of [225Ac]DOTA-HuM195. The % 225Ac that was bound to the HuM195 in the mouse serum was determined as a function of time. IgG bound 225Ac was determined using a Protein A binding assay, HPLC size exclusion chromatography (SEC) analysis of the serum, and a cell based immunoreactivity assay. The results of the Protein A bead assay at 5 time-points from 2.5 to 120 hours, showed that the mean % 225Ac that was bound to the HuM195 was 98.2 to 99.9%, respectively. HPLC SEC analysis revealed that the 225Ac species in the serum was associated with an antibody sized molecule. The immunoreactivity assay of a serum sample indicated that 63% of the 225Ac species was bound to CD33 expressing AL67 cells versus 3% bound to non-specific Daudi cells. In conclusion, 225Ac bound to HuM195 remained associated with the IgG following injection into a mouse over a 5 d period, demonstrating the stability of the drug in vivo .
The in vitro cytotoxicity of 225Ac-antibody constructs that were designed to specifically target HL60 leukemia cells (HuM195 (anti-CD33)); Daudi and Ramos lymphoma cells (B4 (anti-CD19)); MCF7 breast carcinoma cells (trastuzumab (anti-HER2/neu)); LNCaP.FGC prostate carcinoma cells (J591 (anti-PSMA)); and SKOV3 ovarian cancer cells (trastuzumab (anti-HER2/neu)) were examined using very small doses of nanogenerators. The LD50 values of the 225Ac constructs ranged from 0.3 to 74 Bq/mL (0.008 to 2 nCi/mL) and were 2–4 logs lower than activity values for corresponding 213Bi alpha-particle emitting antibodies (see Table I). Controls at low specific activities (target sites were blocked by addition of excess unlabeled ‘cold’ antibody) did not show specific binding of the nanogenerators to target, and were used to evaluate non-specific cytotoxicity. The LD50 values were 10- to 625-fold greater in the blocked controls .
Comparison of the cytotoxicity in vitro of 225Ac- versus 213Bi-antibody constructs.
ED50 was measured by [3H]thymidine uptake assay .
A pharmacokinetic analysis of the 225Ac construct and two of its progeny was performed in vivo by injecting 12 kBq of [225Ac]DOTA-J591 or 12 kBq of [225Ac]DOTA-HuM195 (irrelevant control) i.p. in two groups of male athymic nude mice bearing an i.m. LNCaP tumor xenograft. Approximately, 18 and 21 %ID/g of [225Ac]DOTA-J591 was localized in the tumor at 2 and 3 d, respectively. Tumor samples counted 9 min. after sacrifice/harvest, demonstrated that 221Fr was 88% ± 9% and 213Bi was 89% ± 2% of the 225Ac secular equilibrium levels in the tumor. These results indicate the uptake of the parent IgG construct by the tumor and the retention of the progeny at that location . In toxicity experiments, the MTD in naive 20 g mice was 18.5 kBq (500 nCi) [225Ac]DOTA-IgG. Mice injected with 37 kBq (1000 nCi) of [225Ac]IgG died. Based on these studies, RIT doses were selected that were approximately 40% of MTD .
Therapeutic efficacy of [225Ac]DOTA-J591 was evaluated in an i.m. LNCaP tumor model in vivo. Serum prostate specific antigen (PSA) was utilized in this xenograft model to follow tumor growth. The experimental groups of animals had mean PSA values of 2–5 ng/mL on 10 and 12 d after implantation of tumor. At the time the [225Ac]DOTA-J591 was administered on day 12 or 15, the tumors were characterized histologically as vascularized and encapsulated nodules each comprised of tens of thousands of cells. Animals were sacrificed when tumor area was ≥ 2.5 cm2. In the first RIT study, mice were treated on day 15 post-tumor implantation and received 7.2 kBq [225Ac]DOTA-J591 in a single nontoxic administration. These animals had significantly improved median survival times relative to mice treated with a similar dose of [225Ac]DOTA-B4 irrelevant control antibody mixed with unlabeled specific J591 (dual control) or untreated controls. There was no significant difference in survival times between the dual control-treated animals and untreated controls. The median survival time of untreated growth controls in this model was 33 d. The mean and median pre-therapy PSA values measured on day 12 were not significantly different between the three groups of mice. However, on days 28 and 42, the PSA values of [225Ac]DOTA-J591 treated animals were significantly lower than the PSA values for the dual control-treated animals and untreated controls. There was no significant difference between the dual control-treated animals and untreated controls at either time. Additionally, no acute radiotoxicity was observed .
In a second RIT study, mice were treated on day 12 after LNCaP tumor implantation with a single, non-toxic administration of 7.8 kBq [225Ac]DOTA-J591 which caused tumor regression and significantly improved the median survival times of these mice to 158 d compared to the 63 d in the mice treated on day 15. PSA values decreased from pre-therapy levels in many of the animals following treatment to low and undetectable levels and remained undetected in the 14 of the 39 treated animals which exhibited prolonged survival. These mice survived at least 10 months and had no measureable PSA or evidence of tumor at the time of sacrifice (293 d). Animals treated with unlabeled J591 (0.004 or 0.04 mg) on day 12 post-implantation had no prolongation of median survival (37d and 35d, respectively). The therapeutic efficacy was dependent on antibody specificity, the administration of the 225Ac-generator, and the treatment time after implantation .
In order to determine if other tumor types could be treated with 225Ac-generator constructs, a disseminated human Daudi lymphoma cell mouse model using [225Ac]DOTA-B4 as the therapeutic agent was investigated. SCID mice were treated 1 d after tumor dissemination with a single administration of specific [225Ac]DOTA-B4 (three different dose levels), irrelevant control [225Ac]DOTA-HuM195 (two dose levels), or unlabeled B4. Control mice receiving the irrelevant [225Ac]DOTA-HuM195 had median survival times from xenograft of 43 d (5.6 kBq) and 36 d (1.9 kBq). Mice receiving 0.003 mg unlabeled B4 per mouse had a median survival time of 57 d. The mice receiving a single injection of [225Ac]DOTA-B4 showed dose-related increases in median survival times 165 (6.3 kBq), 137 (4.3 kBq), and 99 d (2.1 kBq), respectively. This dose response of [225Ac]DOTA-B4 was significant and about 40% of mice treated at the highest dose were tumor-free at 300 d and the experiment concluded on day 310 . The images in Figure 2 demonstrate the potency of the [225Ac]DOTA-B4 drug construct against this disseminated lymphoma model.
Bioluminescence imaging (BLI) of two groups of scid mice that were xenografted with Daudi tumor cells transfected with the green fluorescent protein (GFP) and firefly luciferase (FFLuc) genes . Images were taken on day 17 (a) after treatment with [225Ac]DOTA-B4 or (b) untreated growth controls. In the scid model, the GFP+/FFLuc+ Daudi cells developed into macroscopic, disseminated tumors in the bone marrow and spleen as well as in kidneys, liver, lungs, ovaries, and adipose tissue. BLI clearly showed the presence of lymphoma in the untreated mice while no disease was detected in the mice treated with the [225Ac]DOTA-B4. (n.b., the scale bar indicate the value x 1E6 photons/sec/cm2).
The time of treatment from tumor implantation was examined in the second disseminated lymphoma experiment in vivo. Mice that received treatment on day 1, 3, or 6 post tumor implantation with a single administration of [225Ac]DOTA-B4 (6.3 kBq) had similar prolongation of survival relative to untreated growth controls. Mice that received treatment 13 d after tumor dissemination survived > 165 d. Unlabeled B4 was minimally active in mice with median survival of 44 and 40 d for mice treated with 0.002 mg or 0.20 mg, respectively. Untreated growth controls had a median survival time of 28 d. Therefore, in this lymphoma model, while specificity and dose level were important factors in efficacy, the treatment time after tumor dissemination was less relevant up to a time, at which it was then inversely related to activity. The latter phenomenon may be related to differences in the geometry of the alpha emission eradicating single cells or clusters of tumor cells .
Following these initial studies, the same strategy was employed in a RIT experiment in an i.p. mouse model of human SKOV3 ovarian cancer using [225Ac]DOTA-trastuzumab, an anti-HER-2/neu construct . Construct that was administered i.p., had a high tumor uptake, 60 %ID/g at 4 h. Tumor uptake was 3–5-fold higher than liver and spleen. RIT was examined with native trastuzumab and doses of 220, 330 and 450 nCi of [225Ac]DOTA-trastuzumab or [225Ac]DOTA-labeled control antibody at different dosing schedules. Therapy was initiated 9 d after tumor seeding. Groups of untreated control mice and those administered native trastuzumab had median survivals of 33 and 44 d, respectively. Median survival was 52–126 d with [225Ac]DOTA-trastuzumab at various doses and schedules and 48–64 d for [225Ac]DOTA-labeled control IgG. Radiotoxicity occurred with only the highest activity levels administered, but other dose levels were safe. It was concluded that i.p. administration with an internalizing [225Ac]DOTA-labeled anti-HER2/neu antibody significantly extended survival in a mouse model of human ovarian cancer at levels that produce no apparent gross toxicity.
In anticipation of starting human clinical trials, the pharmacokinetics, dosimetry and toxicity of [225Ac]DOTA-HuM195 was investigated in cynomolgus monkeys . The monoclonal antibody, HuM195 (anti-CD33), was the targeting molecule intended for human clinical trials of [225Ac]DOTA-IgG directed against leukemia. In one experiment, two monkeys received a single i.v. dose of the construct at 28 kBq/kg. This dose level was approximately that planned for initial human dose. In another experiment, two animals received a dose escalation schedule of three increasing [225Ac]DOTA-HuM195 doses with a cumulative activity of 377 kBq/kg. Cynomolgus monkeys do not express human CD33 and thus there were no targets for this antibody in this system. The blood half-life of the construct; the ratio of 225Ac:213Bi; the generation of monkey anti-human antibodies (MAHA); haematological indices; serum biochemistries; and clinical observation were the parameters that were measured to evaluate toxicity. Monkeys were euthanized and examined histopathologically when the dose escalation study exhibited toxicity. The blood half-life of [225Ac]DOTA-HuM195 was 12 d and 45% of generated 213Bi daughters were cleared from the blood. MAHA production was not detected. A dose of 28 kBq/kg of 225Ac caused no toxicity at 6 months, whereas a cumulative dose of 377 kBq/kg caused severe toxicity. In the cumulative dosing schedule experiment, single doses of about 37 kBq/kg resulted in no toxicity at six weeks. After 130 kBq/kg was administered, no toxicity was observed for 13 weeks. However, 28 weeks after this second dose administration, mild anemia and increased blood urea nitrogen (BUN) and creatinine were detected indicating renal toxicity. Following administration of an additional 185 kBq/kg, toxicity became clinically apparent. Monkeys were euthanized 13 and 19 weeks after the third dose administration (cumulative dose was 377 kBq/kg). Histopathological evaluation revealed renal tubular damage associated with interstitial fibrosis. In conclusion, the 225Ac nanogenerator constructs may result in renal toxicity and anemia at high doses. The longer blood half-life and the lack of target cell antigens in cynomolgus monkeys may increase toxicity compared to human application. It was concluded that a dose level of 28 kBq/kg could be a safe starting dose in humans, and that hematologic and renal function will require close surveillance during clinical trials.
A model of neuroblastoma meningeal carcinomatosis was treated with an intrathecal (i.t.) administration of [225Ac]DOTA-3F8 construct that targeted ganglioside GD2 . 3F8 is an antibody that specifically binds to ganglioside GD2, overexpressed by many neuroendocrine tumors including neuroblastoma (NB). The [225Ac]DOTA-3F8 construct was prepared and evaluated for radiochemical purity, sterility, immunoreactivity, cytotoxicity in vitro, induction of apoptosis on GD2-positive cells, as well as pharmacologic biodistribution and metabolism of the 225Ac generator and its daughters in a nude mouse xenograft model of NB. Therapeutic efficacy was examined in a nude rat xenograft model of meningeal carcinomatosis and an additional toxicity study in cynomolgus monkeys was performed after i.t. administration. [225Ac]DOTA-3F8 biodistribution in mice showed specific targeting of a s.c. tumor with some redistribution of the 225Ac daughter nuclides mainly from blood to kidneys and to small intestine. In an aggressive meningeal carcinomatosis xenograft nude rat model, i.t. RIT improved survival time two-fold. Increasing the construct specific activity (S.A.) to > 1 MBq/mg improved the therapeutic efficacy relative to lower S.A. preparations. Monkeys injected i.t. with multiple doses of the [225Ac]DOTA-3F8 prepared under clinical manufacturing conditions, did not show any signs of toxicity based on blood chemistry and by complete blood counts or by clinical examination.
Breast cancer spheroids with different HER2/neu expression levels were treated with [225Ac]DOTA-trastuzumab . The breast carcinoma cell lines MCF7, MDA-MB-361, and BT-474 with relative HER2/neu expression of 1:4:18 (determined by flow cytometry) were used. Spheroids of these cell lines were incubated with different concentrations of construct, and spheroid growth was measured by light microscopy over a 50 d period. The activity concentration required to yield a 50% reduction in spheroid volume at 35 d was 18.1, 1.9, and 0.6 kBq/mL (490, 52, 14 nCi/mL) for MCF7, MDA, and BT-474 spheroids, respectively. MCF7 spheroids continued growing, but with a 20–30 d growth delay at 18.5 kBq/mL. MDA-MB-361 spheroid growth was delayed by 30–40 d at 3.7 kBq/mL and at 18.5 kBq/mL, 12 of 12 spheroids disaggregated after 70 d and cells remaining from each spheroid failed to form further colonies. Eight of 10 BT-474 spheroids failed to regrow at a construct concentration of 1.85 kBq/mL. All of the BT-474 spheroids at activity concentrations 3.7 kBq/mL failed to regrow and form colonies. The radiosensitivity of these three cell lines evaluated as spheroids was described as the activity concentration required too reduce the treated-to-untreated spheroid volume ratio to 0.37, denoted DVR37. The external beam radiosensitivity for spheroids of all three cell lines was found to be 2 Gy. After α-particle irradiation using the construct yielded a DVR37 of 1.5, 3.0, and 2.0 kBq/mL for MCF7, MDA-MB-361, and BT-474, respectively.
Targeting the aberrantly formed and angiogenic neovasculature in tumors with an 225Ac-labeled antibody construct has proved very to be effective in prolonging survival and improving subsequent chemotherapy in prostate  and adenocarcinoma  xenograft models. Angiogenic vascular endothelium (VE) expresses the monomeric cadherin (VE-cadhm) epitope on the cell surface that upon dimerizing with another monomeric copy of VE-cadhm on an adjoining cell forms a tight adherens junction between the cells. The antibody E4G10 binds only to VE-cadhm and not the homodimeric form (the binding region is masked in the homodimer), thus conferring specificity for targeting angiogenic and poorly joined VE cells in vivo while not binding to normal VE or tumor. We have demonstrated that [225Ac]DOTA-E4G10 could specifically irradiate carcinoma VE cells as well as bone marrow-derived endothelial progenitors (42) and delay tumor growth. We have also examined [225Ac]DOTA-E4G10 in vascular targeted strategies to treat animal models of glioblastoma multiforme with similar results (preliminary data). Treatment with 1.85 kBq (50 nCi) [225Ac]DOTA-E4G10 on day 3, 5, 7 and 10 after xenotransplant of LNCaP prostate carcinoma cells achieved highly significant inhibition of tumor growth and lower PSA values 22 d after tumor implantation over [225Ac]DOTA-non-specific IgG and vehicle . The lack of binding to tumor cells and to normal vasculature was demonstrated by flow cytometry, by SPECT imaging and by biodistribution studies. Additionally, subsequent bi-weekly administration of paclitaxel for two weeks resulted in further enhancement of the anti-tumor response to survival times of 182 d, compared to [225Ac]DOTA-E4G10 monotherapy (113 d) and to combination of 225Ac-labeled unspecific IgG with paclitaxel (84 d). The authors concluded that targeting the neovasculature with alpha particles was an effective approach to cancer therapy and that sequential therapy with chemotherapy could potentially result in a synergistic effect when temporal administration was carefully planned.
Disrupting and damaging the vascular endothelial architecture associated with tumor tissue is a viable therapeutic strategy. The endothelial vessels in tumor vasculature often do not exhibit the same hierarchy as in normal tissue. Instead, tumor vascular networks are tortuous and have abnormal component and structural composition. Endothelial cells in these tumors are inefficiently joined with holes, gaps and defects; pericytes are loosely associated with vessels or absent; and basement membranes are inefficiently applied relative to typical normal tissues. Alpha particles are charged helium nuclei that travel approximately 50–80 μm, the dimensions of a typical vessel within a tumor. In addition, individual α-particles are able to kill a target cell due to their deposition of 5–8 MeV in an ionizing track that is several cell diameters in length. Alpha particles are very potent cytotoxic agents in this close vicinity but will largely spare normal tissue; it is this characteristic that offers clear advantages to other known forms of targeted radiation (i.e., β−- or auger-emitters) or stereotactic external beam therapy as a means of effecting selective cell kill.
In the colon adenocarcinoma model system, a novel mechanism of achieving vascular normalization and improved chemotherapy with a cytotoxic anti-vascular agent was described . Selective cytotoxicity was directed towards the tumor neovasculature using short-ranged alpha particles targeted to VE-cadhm. Immunofluorescence and immunohistochemical studies showed that the alpha-targeted vasculature of tumors was largely depleted, and that the remaining vessels appeared more mature as substantiated by accompanying morphological changes and increased pericyte density and coverage. Tumor accumulation and micro-distribution studies with radioactive and fluorescent small molecule drugs showed better uptake and more homogenous distribution of the drugs within [225Ac]DOTA-E4G10 treated tumors (versus controls), thus explaining the enhanced therapeutic response. The results showed not only that 225Ac treatment lead to ablation and remodeling of the tumor vasculature, but also suggested that the resulting vessel normalization improved tumor delivery of small molecules.
A metastatic breast cancer model was investigated and compared the therapeutic efficacy 225Ac-, 213Bi-, and 90Y-labeled anti-rat HER-2/neu monoclonal antibody (7.16.4) as the therapeutic agents. A single 400 nCi administration of [225Ac]-7.16.4 completely eradicated breast cancer lung micrometastases in 67% of HER-2/neu transgenic mice and resulted in long-term survival of these mice for up to 1 y. Treatment with 225Ac-7.16.4 was significantly more effective than 120 μCi of [213Bi]-7.16.4 (median survival, 61 d) and 120 μCi of [90Y]-7.16.4 (median survival, 50 d) as well as untreated control (median survival, 41 d). Dosimetric analysis of the 225Ac-treated mice demonstrated that the metastases received a total dose of 9.6 Gy, significantly greater than the 2.0 Gy from 213Bi or 2.4 Gy from 90Y. Biodistribution studies revealed that 225Ac progeny accumulated in kidneys and probably contributed to the long-term renal toxicity observed in surviving mice. These data suggested that an 225Ac-labeled anti-HER-2/neu monoclonal antibody construct could prolong survival in HER-2/neu-positive metastatic breast cancer patients .
A property of such small molecules that limits their application was that after the dose was administered, a relatively high distribution volume was quickly achieved. This occurs in a time that was similar to the radioactive decay of these nuclides. Therefore, the rapid elimination of these small molecules was not advantageous when used with short half-lived radionuclides. However, 225Ac might be an interesting candidate for small molecules in applications where a high tumor accumulation was possible. Peptide receptor radionuclide therapy (PRRT) was evaluated preclinically with 225Ac-labeled 1,4,7,10-tetra-azacylododecane N,N′,N″,N‴-J-tetraacetic acid-Tyr3-octreotide (DOTATOC) against pancreatic neuroendocrine tumor in a xenograft model . Doses of [225Ac]-DOTATOC up to 20 kBq (540 nCi) were not toxic in mice, while activities greater than 30 kBq (811 nCi) induced tubular necrosis. Biodistribution studies revealed that [225Ac]-DOTATOC effectively localized in the neuroendocrine tumors with some kidney accumulation. Doses of 12–20 kBq (324–540 nCi) of [225Ac]-DOTATOC effectively controlled neuroendocrine tumor growth and showed improved efficacy compared with [177Lu]-DOTATOC and DOTATOC controls.
Innovations in pharmaceutical design are envisioned that seek to improve the potency and specificity of conventional agents by integrating nanomaterials into the drug construct blueprint. For example, novel synthetic nanostructures based on molecules consisting of biologics, radionuclides and carbon nanotubes will have emergent anti-cancer properties because of amplified targeting, binding, imaging, therapeutic and novel pharmacokinetic characteristics. These nanomaterials should therefore exhibit improved potency, specificity, and efficacy relative to conventional agents. Soluble carbon nanotube constructs, containing multiple copies of covalently attached antibodies, and chelated radiometals, have been shown to target lymphoma and adenocarcinoma in imaging and therapeutic studies in animal models of human disease [45,46]. Prototypes of these nanoconstructs rapidly cleared the blood (t1/2 < 1 h), had minimal accumulation in liver, spleen and kidney, and were rapidly cleared intact into the urine (t1/2 ~ 6 min.) by glomerular filtration [47,48].
The production and radio-chromatographic properties of a metallofullerene encapsulated 225Ac were reported. Chromatographic results suggested a 3+ oxidation state of Ac in the Ac@C82 complex and the presence of metallofullerene isomers with properties similar to La@C82 [49,50].
Pegylated phosphatidylcholine-cholesterol liposomes with encapsulated 225Ac (passively entrapped) were developed to retain the potentially toxic daughters at the tumor site. More than two 225Ac atoms were successfully entrapped per liposome and more than 88% of the activity was retained over 30 d. The size of the liposomal structures required to contain the daughters makes this approach ideally suited for locoregional therapy (e.g., i.p., intrahepatic artery, or intrathecal) .
Improved daughter retention was realized using multivesicular liposomes (MUVEL). MUVELs are large pegylated liposomes with the 225Ac entrapped within smaller lipid-vesicles. This strategy provides confinement of entrapped 225Ac within the region of the liposomal core, away from the outer liposomal membrane. These MUVELs yielded 98% retention of 225Ac and 18% retention of the last daughter 213Bi for 30 d. MUVELs were then conjugated to trastuzumab and exhibited robust binding and internalization by ovarian carcinoma cells. [225Ac]-MUVELs were administered i.p. to animals with disseminated disease and significant tumor uptake of 225Ac and its daughters was detected .
Targeting cancer cells that express low levels of antigens is an issue with strategies using radiolabeled carriers with low specific activities. Antibodies may not deliver enough α-emitters to the targeted cancer cells to result in killing, but liposomes with conjugated with targeting antibodies were loaded with high levels of 225Ac to overcome the limitations of low specific activity. As expected these were demonstrated to be therapeutically useful against tumor cells having a low antigen density .
Large (approximately 600 nm in diameter) pegylated liposomes conjugated with trastuzumab resulted in swift, specific binding to cancer cells in vitro, followed by cellular internalization. After i.p. administration, these liposomes again exhibited rapid, specific binding to tumors. The large liposomes were cleared slowly from the i.p. cavity, exhibited an increased uptake by the spleen relative to the liver, and specifically targeted tumor. The findings suggested that large targeted liposomes administered i.p. could be a potent drug-delivery strategy for locoregional therapy of i.p. micrometastatic tumors .
The extreme potency of 225Ac necessitates the small doses administered for preclinical and clinical therapeutic studies; as a consequence, planar SPECT imaging of 225Ac or its 213Bi daughter is not feasible . The development of a novel optical imaging technique was recently reported the used the inherent optical emissions from radionuclide decay for Cerenkov luminescence imaging (CLI) of tumors in vivo. The results correlated with those obtained from concomitant immuno-PET studies using analogous β+-emitting immunoconstructs. Phantom studies confirmed that Cerenkov radiation was observed from a range of positron-, beta-, and alpha-emitting radionuclides using standard optical imaging devices. The change in light emission intensity versus time linearly correlated with radionuclide decay, activity concentration, and the measured PET signal (%ID/g). The value of CLI lies in its ability to image radionuclides that do not emit either β+ or γ-rays and are unsuitable for use with current nuclear imaging modalities. Optical imaging of Cerenkov radiation emission shows excellent promise as a potential new imaging modality for the rapid, high-throughput screening of radiopharmaceuticals .
Further efforts to understand the pharmacokinetics of the 225Ac progeny and the elucidation of methods to control their biodistribution and elimination have been described in order to have a better idea of the consequences of therapy in vivo . The combination of stable DOTA chelation of 225Ac, efficient targeting of cell specific epitopes, and internalization of the constructs led to a potent and effective therapeutic nanogenerator strategy no systemic toxicity observed. The internalization of the 225Ac-labeled construct coupled with the stable DOTA chelation during targeting proved useful in harnessing the cytotoxic potential of the daughters and mitigating their effects.
Groups of naïve mice were administered (i.v.) 500 nCi of [225Ac]DOTA-IgG. Metallic progeny chelation was effected with either 2,3-dimercapto-1-propanesulfonic acid (DMPS) or meso-2,3-dimercaptosuccinic acid (DMSA). Both chelating agents significantly reduced the renal uptake of 213Bi; however, DMPS was more effective than DMSA. Diuresis with furosemide or chlorothiazide (CTZ) treatment significantly reduced the renal 213Bi and 221Fr activities. The combination of DMPS with either CTZ or furosemide further reduced renal 213Bi activity. Competitive antagonism with ’cold’ bismuth subnitrate only moderately reduced the renal uptake of 213Bi. In studies with tumor-bearing mice, the tumor ‘sink’ significantly prevented the renal 213Bi accumulation as the daughter was presumably in the tumor. The tumor sink-effect combined with DMPS treatment further reduced renal 213Bi activity. The results indicated that metal chelation, diuresis with furosemide or CTZ, and competitive metal blockade could serve as adjuvant therapies to modify the potential nephrotoxicity of 225Ac progeny. Furthermore, internalization of the parent construct to tumor decreased non-specific organ uptake of 213Bi .
A pharmacokinetic study of the short-lived daughter nuclide 221Fr was performed in naïve mice . The majority of the progeny biodistribution studies (described above) focused on 213Bi, so a source of 221Fr was developed for comparison. An 225Ac/221Fr generator was designed and constructed. Briefly, a DOTA-biotin construct  was labeled with 225Ac and reacted with an immobilized avidin column. The generator was eluted with normal sterile saline yielding predominantly 221Fr activity. The 221Fr biodistribution study only harvested and counted blood and kidneys because of the short nuclide half-life. The %ID/g of 221Fr was 52.3 ± 8.4 and 5.4 ± 0.3 in the kidneys and blood of these animals (n = 3), respectively.
Some of the systemically released 225Ac progeny accumulated in the kidneys. The ensuing renal tubulointerstital changes were investigated in order to elucidate the radiobiological effects of the daughters in the kidney . Toxicological and histopathological evaluations of nonhuman primates (cynomolgus monkeys) which received a cumulative activity of 377 kBq/kg in a dose escalation schedule of [225Ac]DOTA-HuM195, revealed mainly renal tubular damage associated with interstitial fibrosis . Upto this time, the mechanism of radiation nephropathy that resulted from targeted radionuclide therapies was poorly understood.
Naïve mice were administered 350 nCi of [225Ac]DOTA-HuM195 and the subsequent functional and morphological changes in their kidneys were assessed longitudinally. Renal irradiation from free, radioactive 225Ac progeny led to time-dependant reduction in renal function manifested as tissue pallor and an increased blood urea nitrogen titer. Corresponding histopathological changes were observed in the kidneys. Glomerular and tubular cell nuclear pleomorphism, karyorrhexis, tubular cell injury and lysis were observed as early as 10 weeks. Progressive thinning of the cortex due to widespread tubulolysis, collapsed tubules, glomerular crowding, decrease in glomerular cellularity and interstitial inflammation and an elevated juxtaglomerular index were noted at 5–7 weeks post-treatment. By 35 – 40 weeks, regeneration of simplified tubules with tubular atrophy and loss and focal interstitial fibrosis had occurred. A lower juxtaglomerular cell index with focal cytoplasmic vacuolization was observed and suggested increased degranulation. Increased tubular and interstitial TGF-β1expression and a corresponding increase in the extracellular matrix deposition was noticed only at 40 weeks post-injection. These findings suggest that internally delivered α-particle radiation-induced loss of tubular epithelial cells and triggered a procession of adaptive changes that resulted in progressive morphological damage accompanied by a loss of renal function.
Radiation nephropathy that followed internal alpha particle irradiation of kidneys was ameliorated by pharmacologically modifying the functional and morphological changes in mouse kidneys following injection of [225Ac]DOTA-HuM195 using several different agents . The 350 nCi dosage yielded a 27.6 Gy dose to the kidneys. Mice were randomized to receive captopril (ACE inhibitor), L-158,809 (Angiotensin II receptor-1 blocker), spironolactone (aldosterone receptor antagonist) or a placebo control. Forty weeks after [225Ac]DOTA-HuM195 injection, placebo-control mice showed significant increase in BUN, dilated Bowman spaces and tubulolysis with basement membrane thickening. Spironolactone treatment significantly prevented the development of adverse histopathological and functional changes vs. placebo controls. The Angiotensin II receptor-1 blocker offered moderate protection. Captopril treatment accentuated the functional and histopathological damage. In conclusion, low-dose spironolactone, and to a lesser extent, angiotensin receptor-1 blockade, offered renal protection in a mouse model of internal alpha particle irradiation.
Actinium-225 and its progeny present unique radiobiological and dosimetric challenges to the medical physicist because of the multiple progeny, the diverse progeny chemical characteristics and pharmacokinetic profiles, and the short-range, high LET α-particle emissions. Recently, MIRD Pamphlet No. 22 was published to address these issues for 225Ac and other α-particle emitters [62,63]. A plan was developed for estimating absorbed dose to organs following the administration of radionuclides with multiple progeny in order to model the dosimetry of 225Ac its daughters . This dosimetric evaluation of α-particle emitters required that all decays, including those of unstable intermediates be included in the calculation. These calculations were complicated by the differential biodistributions of each of the progeny due to their varied periodic properties. The formalism that was presented accounted for the known pharmacokinetic profiles of the daughters and the effective biodistribution focused on the site at which the 225Ac decayed. The cumulative decays of a daughter present in a particular tissue were estimated using a probability matrix which described the likelihood of daughter decay in a particular tissue as a function of the decay site of the parent.
Cellular dose conversion factors (DCF) were calculated employing the dose contributions of several progeny at the site of 225Ac decay that were made dependent on a threshold time parameter . This enabled an estimate of the fraction of daughter decays expected at the site of parent decay. Previously tabulated S values (cell-surface to nucleus and cell-surface to cell) for each daughter were then scaled by this fraction and the sum over all progeny was performed to yield a cut-off time-dependent set of corresponding DCF values. These DCF values for the absorbed dose to the nuclear or cellular volume arising from cell-surface decays were presented as a function of the cut-off time for several different representative sets of cellular and nuclear dimensions. In contrast to the cellular S values that accounted only for the 225Ac decay, these cellular DCF values made it now possible include the contribution of progeny decays in cellular α-particle emitter dose calculations.
A theoretical estimate of the absorbed dose to key organs arising from the use of radionuclides with multiple unstable progeny was made using three sophisticated model features applied to different sized tumor masses and carrier specifications . Each of the 225Ac progeny of has its own biodistribution profile and half-life, therefore, including their contributions would yield a more accurate prediction of absorbed dose and potential toxicity. The first model restricted the transport to a function that yielded either the place of origin or the place(s) of biodistribution depending on the half-life of the parent radionuclide. The second model included the transient time in the bloodstream and the third model incorporated additional binding at or within the tumor. (Note that the second model allowed for radionuclide decay and additional daughter production while transiting from one location to the next and that the third model relaxed the constraint that the residence time within the tumor was solely based on the half-life of the parent.) Calculations simulated were both a rapidly accessible small (0.1 g) tumor and a large (10 g) solid tumor. In addition, the effects of varying the carrier molecule purity and mass amount, as well as tumor cell antigen saturation were examined. The results indicated that there was a distinct advantage in using a parent radionuclide such as 225Ac, having a 10 d half-life and yielding 4 alpha particles per decay. Lower normal tissue doses resulted for a given tumor dose in comparison to those radionuclides yielding fewer alpha particles.
[225Ac]-lintuzumab (HuM195) is in use in an ongoing Phase I clinical trial at Memorial Sloan-Kettering Cancer Center. At press, the twelfth patient had been treated. Patient accrual continues. This Phase I clinical trial with [225Ac]-lintuzumab resulted from a period of clinical investigation of CD33-targeted therapy in patients with AML, mainly relapsed or refractory. Early studies were conducted with the murine form, M195, labeled with 131I in patients with minimal residual disease (50 – 70 mCi/m2) or to intensify therapy prior to bone marrow transplant (BMT) (120–230 mCi/m2). However, the detection of human anti-mouse antibodies in a fraction of patients precludes additional M195 treatments in the further course of the disease . The humanized form of this antibody, HuM195, demonstrated a lack of immunogenicity and an increased affinity to CD33. Intensification of therapy prior to BMT might be achieved with the longer ranged beta-emitter yttrium-90 and further clinical studies where conducted with 90Y labeled HuM195 for myeloablation . In contrast, in non-myeloablative regimens, CD33 negative stem cells have to be spared from the non-specific cross radiation. This led to the strategy that employed the alpha-emitting nuclide 213Bi labeled to HuM195. In a Phase I clinical trial, the anti-leukemic effect of [213Bi]-HuM195 was demonstrated in patients with relatively high tumor burden . Because of the large tumor burden in AML, the Phase I/II study used a regimen wherein Cytarabine was given prior to the [213Bi]-HuM195 to effect some cytoreduction before alpha-particle therapy. This led to the first clinical trial using the longer-lived and more potent 225Ac in humans using the [225Ac]DOTA-HuM195 construct .
A remarkable therapeutic efficacy has been demonstrated with 225Ac-prostate-specific membrane antigen (PSMA)-617 in heavily pre-treated metastatic castration-resistant prostate cancer (mCRPC) patients. We report our experience with 225Ac-PSMA-617 therapy in chemotherapy-naïve patients with advanced metastatic prostate carcinoma.
Seventeen patients with advanced prostate cancer were selected for treatment with 225Ac-PSMA-617 in 2-month intervals, with initial activity of 8 MBq, then de-escalation to 7 MBq, 6 MBq or 4 MBq in cases of good response. In one patient, activity was escalated to 13 MBq in the third cycle. Fourteen patients had three treatment cycles administered, while in three patients treatment was discontinued after two cycles due to good response. Six out of 17 patients received additional treatments after the third cycle. Prostate-specific antigen (PSA) was measured every 4 weeks for PSA response assessment. 68Ga-PSMA-PET/CT was used for functional response assessment before each subsequent treatment cycle. Serial full blood count, renal function test, and liver function were obtained to determine treatment-related side effects.
Good antitumor activity assessed by serum PSA level and 68Ga-PSMA-PET/CT was seen in 16/17 patients. In 14/17 patients, PSA decline ≥90% was seen after treatment, including seven patients with undetectable serum PSA following two (2/7) or three cycles (5/7) cycles of 225Ac-PSMA-617. Fifteen of 17 patients had a > 50% decline in lesions avidity for tracer on 68Ga-PSMA-PET/CT including 11 patients with complete resolution (PET-negative and either stable sclerosis on CT for bone or resolution of lymph node metastases) of all metastatic lesions. Grade 1/2 xerostomia was seen in all patients, and none was severe enough to lead to discontinuation of treatment. One patient had with extensive bone marrow metastases and a background anemia developed a grade 3 anemia while another patient with solitary kidney and pre-treatment grade 3 renal failure developed grade 4 renal toxicity following treatment. The group presented with significant palliation of bone pain and reduced toxicity to salivary glands due to de-escalation.
225Ac-PSMA-617 RLT of chemotherapy-naïve patients with advanced metastatic prostate carcinoma led to a ≥ 90% decline in serum PSA in 82% of patients including 41% of patients with undetectable serum PSA who remained in remission 12 months after therapy. The remarkable therapeutic efficacy reported in this study could be achieved with reduced toxicity to salivary glands due to de-escalation of administered activities in subsequent treatment cycles. This necessitates further exploration for informing clinical practice and clinical trial design.
The information comes from:https://www.ncbi.nlm.nih.gov/pubmed/30232539
Brachytherapy employing iodine-125 seeds is an established treatment for low-risk prostate cancers. Post-implant dosimetry (PID) is an important tool for identifying suboptimal implants. The aim of this work was to improve suboptimal implants by a subsequent iodine-125 seed top-up (reimplantation), based on the PID results.
Of 255 patients treated between 2009 and 2012, 6 were identified as having received suboptimal implants and were scheduled for seed top-up. Needle configurations and the number of top-up seeds were determined based on post-implant CT images as well as a reimplantation treatment plan. An average of 14 seeds per patient were implanted during each top-up. Dosimetric outcome was assessed via target parameters and doses received by organs at risk.
All six patients had a successful top-up, with a 67% increase in the mean dose delivered to 90% of the prostate volume and a 40% increase in the volume that receives 100% of the prescribed dose. However, the final dosimetric assessment was based on the same seed activity, as the planning system does not account for the decay of the initially implanted seeds. Although physical dosimetry is not influenced by different seed activities (doses are calculated to infinity), the radiobiological implications might be slightly different from the situation when optimal implantation is achieved with one treatment only.
Seed reimplantation in suboptimal prostate implants is feasible and leads to successful clinical outcomes.
Suboptimal prostate implants can occur for various reasons. This work shows that seed reimplantation as salvage therapy can lead to an optimal dosimetric outcome with manageable normal tissue effects.
Low dose rate (LDR) brachytherapy employing radioactive seeds is a well-established treatment for low-risk prostate cancers. In our centre, implants are conducted with iodine-125 seeds (Oncura RAPID Strand, model 6711; Oncura Inc., Arlington Heights, IL) with an average seed activity of 0.395 mCi to deliver a prescribed dose of 145 Gy (to >98% of the prostate). Treatment planning and post-implant dosimetry (PID) are completed using SPOT-PRO™ v. 3.1 (Nucletron, Utrecht, Netherlands) software based on transrectal ultrasound images (for treatment planning) and CT images (for PID).
The most commonly reported parameters that are indicative of the dosimetric quality of the implant are D90 (the dose delivered to 90% of the prostate volume) and V100 (the volume that receives 100% of the prescribed dose). Several studies showed a link between the quality of implants and the biochemical outcome [1–3]. Therefore, to minimise the risk of recurrence, it is recommended to achieve a post-implant D90>140 Gy and V100>90% .
PID is an important quantitative tool for the assessment of LDR implants; therefore, it is recommended as a routine procedure by several professional organisations [2–5]. Besides evaluating the overall quality of the implant, PID can assist in the dosimetric assessment of the organs at risk (OARs). Although the dosimetry of OARs cannot be adjusted if overdosed, the radiation oncologist can have a closer follow-up of those patients at risk of developing normal tissue sequelae.
Another role of PID is to identify suboptimal implants that can arise owing to organ movement during the procedure, geographical misses of seeds or technical equipment errors. Despite all the efforts and experience of the brachytherapy team, suboptimal implants do occur and they have to be dealt with. Although several centres encounter such events, there is a lack of guidelines or even indications as to how to proceed to improve the final outcome. The major challenge is perhaps the planning, which cannot be done in a conventional way, i.e. based on the ultrasound study of the transrectal volume, owing to lack of previously implanted seed visibility . Therefore, post-implant CT images are the most convenient to use for this task, as the original seeds can be seen and extra seeds can be added to cover the underdosed areas of the prostate.
The aim of this work was to present our experience with iodine-125 seed reimplantation (top-up) in a cohort of six patients whose initial implant was suboptimal as identified by PID. The technical and dosimetric challenges of seed top-up implants are investigated.
Of 255 patients treated with iodine-125 permanent seed brachytherapy between January 2009 and July 2012 at the Royal Adelaide Hospital, SA, 6 were identified as having received suboptimal implants and were scheduled for seed top-up. The suboptimal implants were attributed to equipment/technical problems (two cases), patient movement during implantation (one case) and uncommonly large oedema (one case); however, for two patients, the causes of underdosage were unidentified. Therefore, the cold spots were random from patient to patient, although underdosage of the base was more common than the involvement of other prostate regions.
The decision to undergo reimplantation was based on the PID results using ultrasound–CT image coregistration, which our centre’s standard technique (the methods are described in detail in a previous report ). However, in all situations, the main dosimetric parameters were independently confirmed by a second medical physicist, and in some cases, when the result was uncertain, an alternative PID technique was also used, such as CT-delineation-based PID or MR–CT image coregistration-based PID. As with all our PID data, the results were reviewed by the patient’s radiation oncologist, who, after following the PID prostate D90 criteria given below, decided whether the patient should be scheduled for a seed top-up or not.
At our centre, we categorise the quality of the implants according to the following PID D90 values:
D90>145 Gy—good implant [as recommended by Groupe Européen de Curiethérapie–European Society for Radiotherapy & Oncology (GEC-ESTRO) guidelines] 
130 Gy<D90<145 Gy—adequate by local definition (within 10% of the prescribed dose).
110 Gy<D90<130 Gy—potentially acceptable, subject to re-assessment of PID with another PID technique and evaluation of cold spots. If cold spots are identified in areas that were not positive on transrectal ultrasound biopsy, we consider the implant acceptable. If the implant is subsequently deemed inadequate, the patient is scheduled for a seed top-up procedure.
D90<110 Gy—the implant is inadequate [1,5]. The patient is scheduled for a seed top-up procedure.
For those patients identified as having an inadequate implant, a top-up treatment plan is generated. This is designed using the initial PID ultrasound–CT images as the reference to determine the location and the number of seeds required to improve the dosimetry.
The prostate target (hereafter referred to as the target) is defined as the clinical target volume, and acceptable target volumes for initial implants are between 15 and 55 cm3. A planning target volume (PTV) is also defined during preplanning but is not considered in this report, as its dosimetry trends reflect those of the target values.
This is perhaps the most demanding step in the reimplantation process owing to the limitations of the treatment planning system (as will be discussed later). Some explanations of our PID technique are first required to help understand the seed top-up planning method.
The PID CT images are scanned at 2-mm slice thickness and 2-mm spacing. As described in the study by Marcu et al , the PID CT images are coregistered (matched) to the pre-implant ultrasound images, with the image fusion module of the SPOT-PRO treatment planning system, by manually shifting and rotating the overlaid image sets until anatomical references are aligned. Once image sets are matched, the target and PTV from the ultrasound images are copied onto the CT images; the rectum and urethra are drawn directly onto the CT images. Subsequent analysis of the PID is performed on the CT images with the copied and drawn contours. Following computerised seed identification, the PID is assessed.
Analysis of the PID dose–volume histogram (DVH) data and visual dose coverage identifies the “cold spots” (areas of target receiving less than the prescription dose) and potential areas that require a top-up of seeds. The two main problems that arise with relating the planned positioning of the top-up seeds to the actual desired positions for implant are: (1) the orientation of the CT images and (2) the implant template grid pattern.
(1) Orientation of the CT images: as the CT is performed with the patient supine and the ultrasound is performed with the patient in the lithotomy position, the orientation of the prostate and local anatomy is different in each setting. Although this can be overcome for the purposes of matching two data sets together for copying contours, it presents a problem for determining the seed/needle locations when viewing the CT data set in the planning module. If we assume that the ultrasound data set is the normal orientation, then the CT data set is rotated (approximately 20–40°) about the patient's left/right axis during the matching process; however, when subsequently viewing the CT images in the post-implant planning module, the CT images are in their normal scanned orientation. This means that the needle tracks are not in the horizontal plane, when viewed laterally, but angled anteriorly in the inferior to superior direction (Figure 1). This therefore causes problems with relating possible top-up seed locations (as determined from the CT images) to their equivalent positions when viewed on the ultrasound.
Right lateral views of post-implant CT images at various positions through the prostate. The superimposed manually added lines indicate the needle tracks for the visible seeds. Also visible are the target and planning target volume contours. The central image shows the catheter balloon with contrast and a part of the urethra contour.
The following paragraph describes the method implemented in our department using the SPOT-PRO software. Nevertheless, the essentials of the process could be easily adapted for use with other brachytherapy treatment planning systems. Firstly, we scroll through the lateral CT images and determine the angle of the implanted needles by visualising the seed positions. As the seeds are implanted in the Rapid Strand form, we have confidence in identifying their corresponding needle tracks. Figure 1 shows some superimposed lines that help to illustrate this. Once the angle is determined, we then adjust the CT data set such that we can visualise the transverse CT images that are perpendicular to the angle of the needle tracks.
To achieve this, we use the three-dimensional cube facility within the SPOT-PRO software. Figure 2 demonstrates the steps involved in changing the transverse views from the original to those that are perpendicular to the angle of the needle tracks. The CT needle tracks would now be horizontal if viewed laterally. When we now scroll through the angle-corrected transverse views, they are in the same orientation as the transverse views in the original treatment plan of the ultrasound images. Therefore, determining grid positions for the top-up seeds/needles is achievable.
Demonstration of the steps involved in changing transverse CT images from the normal axial orientation to that which is perpendicular to the needle tracks. Starting from top left in clockwise rotation: a transverse image through the prostate is established within the three-dimensional (3D) cube image set, next a cut plane is created through the image set at an angle perpendicular to the angle of the needle tracks (as shown in Figure 1 and indicated by the protruding line), then the 3D cube is rotated so that the angle-corrected transverse slice is vertical; this is followed by a further rotation to visualise the angle-corrected transverse slice in the conventional way.
(2) Implant template grid pattern: unfortunately, the treatment planning system does not show the implant template grid pattern on the CT images, so mapping out the needle locations according to this grid can be difficult. However, after following the steps above to orientate the post-implant CT images, the grid positions can be determined by simultaneously viewing the CT images and the ultrasound images by using the target contour and seeds as the common reference guide. In the case of the SPOT-PRO planning system, this cannot be done on screen, so we print out a selection of transverse ultrasound images with the template visible and scroll through the CT images on screen. In addition, the base and apex of the target can be identified and compared, so that the depth of needle insertion (for seed deployment) can be determined.
The top-up seeds are then added to the plan (with known potential grid positions) until the cold spots are sufficiently covered with target and satisfactory OAR DVH values. The patients are notified of the treatment outcome and advised that they should undergo a reimplantation procedure. The procedure is guided by transrectal ultrasound and seed deployment is checked on fluoroscopic images, where the previously implanted seeds are also visible. Fluoroscopic guidance is particularly useful in those situations where the underdosage occurs in the base of the prostate, and the new baseline enables a good dosimetric coverage of that area. Dosimetric outcome of the top-up implant is assessed, the following day, via a second post-implant CT scan, coregistered to the original pre-implant ultrasound images as done previously with the initial PID CT. Target parameters (D90 and V100) and doses received by OAR, i.e. urethra [the dose delivered to 10% of the prostatic urethra (D10) and the dose delivered to 30% of the prostatic urethra (D30)] and rectum [the dose to 0.1 cm3 of the rectum (D0.1 cm3) and the dose to 2 cm3 of the rectum (D2 cm3)], are assessed and are reported later.
The number of seeds needed for the initial implant as well as the extra number of seeds used for top-up for all six patients is listed in Table 1. An average of 14 seeds per patient was implanted during each top-up. Table 1 also lists the time interval between the implant and the top-up day. Depending on the waiting list and also the patient’s availability, the top-up was scheduled, on average, within 3 months after the implant.
Seed top-up data
Individual charts showing the target implant dosimetry and the dosimetry after the top-up are represented in Figures 3 and and4.4. The improvement in dosimetry is clearly observed.
Dose delivered to 90% of the prostate volume (D90) target values before and after the top-up individually shown for the six treated patients.
Volume receiving 100% of the prescribed dose (V100) target values before and after the top-up individually shown for the six treated patients.
Figures 3 and and44 illustrate that, if the cold spots had been covered from the initial implant, the main dosimetric parameters D90 and V100 would have significantly increased. Therefore, although the dose to target after the top-up leads to an average D90 of 150 Gy (which represents a 67% increase when compared with the implant dosimetry), the volume to be covered by 100% of the prescribed dose (V100) increased by 40%, reaching an average of 92% for the six studied patients.
All patients but one had D90>140 Gy and V100>90%. Patient 3 had an initial D90=43.5 Gy, increasing to D90=124 Gy after reimplantation. Although the dosimetry is still below the recommended limit, the diseased areas, as proven by biopsies, were all well implanted and the corresponding cold spots were adequately covered. Since the urethral dose was already high before the reimplantation (D10=203 Gy and D30=185 Gy), after the addition of top-up seeds the urethral dose exceeded the recommended safety limits (D10=243 Gy and D30=221 Gy). Thus, to avoid increasing the risk for toxicities, the plan had to be compromised.
The dosimetric evaluation of the OARs was undertaken using the GEC-ESTRO recommendations whereby the urethral primary dosimetric parameter, D10, should be <150% of the prescribed dose and the secondary parameter, D30, should be <130% of the prescribed dose .
The urethral dose related to D10 reached values above the recommended limit for two patients, while according to the secondary dosimetric parameter, D30, three patients received higher doses than recommended limits, as a compromise for a better tumour dosimetry (Table 2).
Post-top-up dosimetry for target and organs at risk
D10, dose delivered to 10% of the prostatic urethra; D30, dose delivered to 30% of the prostatic urethra; D90, dose delivered to 90% of the prostatic urethra; D0.1 cm3, dose to 0.1 cm3 of the rectum; D2 cm3, dose delivered to 2 cm3 of the rectum; V100, volume receiving 100% of the prescribed dose.
Similarly to the prostatic urethra, the dosimetric requirements for the rectum were established based on the GEC-ESTRO recommendations . Therefore, D2 cm3<145 Gy was used as a primary parameter and D0.1 cm3<200 Gy was used as a secondary parameter. One patient had D2 cm3=169 Gy after top-up, whereas, for all the others, the primary dosimetric parameter for the rectum was kept below the 145-Gy limit (Table 2). When assessing the secondary dosimetric indicator, it was noticed that four out of six patients received D0.1 cm3 above the recommended 200 Gy (Table 2). However, two of these patients already had elevated levels before the top-up owing to geographic misplacements of seeds during the original implantation. These patients underwent a closer follow-up to ensure minimum discomfort owing to possible treatment-related side effects. However, no rectal morbidities have been reported so far.
All six patients had a successful top-up, with a 67% increase in the mean D90 value and a 40% increase in V100. Unavoidably, doses to OARs have increased as well, in some cases exceeding the recommended limits. It should be noted that the final dosimetric assessment (implant+top-up) was based on the same seed activity (0.395 mCi), as the planning system does not account for the decay of the initially implanted seeds. The DVH values obtained post top-up are presented in Table 2 both for the target and for the OARs. Table 2 also shows the recommended limits for these structures. It should be noted that, through the top-up procedure, the dose to the target was increased to enable good tumour control, while as a compromise, for the OARs, some values were marginally higher than the recommended limits.
There are two main challenges involving the top-up procedure and its outcome: one covers the physical aspect of seed dosimetry and the other relates to the radiobiological implications of the top-up.
Some planning systems, such as the one used in the current study, do not allow treatment plans with different seed activities related to different implant dates. This means that although on the implant day the average seed activity was 0.395 mCi, by the time of the top-up implant these seeds have decayed by an amount which depends on the time gap between the implant and the top-up day. However, the top-up was conducted using 0.395 mCi seeds, as for the implant; thus, for dose calculation SPOT-PRO assumes that all seeds were implanted on the same date and with the same activity. Despite this discrepancy between the two implanted seed batches used in the dose calculation, any potential differences in physical effect owing to the dose delivered is probably not important for iodine-125 seeds of the same strength but implanted at different times (provided that top-up is given within a couple of isotope half-lives), as they would yield the same cumulative (physical) dose after 2 years (knowing that 80% of the total dose is delivered in about 140 days), to within <0.1%. In our case, the SPOT-PRO dose calculation algorithm uses 10 half-lives to determine the cumulative dose, so any discrepancy in the dose displayed on the plan relating to the different seed activities or dates is insignificant (again provided the top-up is given within 2 isotope half-lives).
Radiobiological effects could be influenced by the time difference between the two implant dates as the 145-Gy prescribed dose is delivered in two separate sessions, thus over a longer time period than initially planned (Figure 5, in which the dose rate over time is calculated considering that the initial dose rate for iodine-125 is about 7 cGy h–1 ). Although patient 1 had his top-up 21 days after implant, thus well within the half-life of iodine-125, and when the dose rate was 5.5 cGy h–1, patient 3 underwent the seed boost 141 days after the implant (owing to logistics problems), at a time when the initially implanted seeds decayed with a dose rate of 1.35 cGy h–1, delivering the remaining 20% of the prescribed dose.
Variation of the initial dose rate of a typical iodine-125 implant with time between implant and top-up.
Nevertheless, considering that permanent seed implants are conducted on patients with low-risk and low-grade prostate cancer, the time gap between the implant and the top-up might not induce any radiobiological events that differ from a single procedure.
Our study was based on the dosimetric evaluation of post-implant CT images registered to ultrasound images. Although this image coregistration was the most convenient in our situation, we acknowledge the fact that fused post-implant CT/MR images registered to ultrasound images are most likely to be the optimum [8,9].
There are limited studies in the literature recommending an external beam radiotherapy (EBRT) boost for suboptimal implants. One such study suggests salvage EBRT using a multimodality dose summation between intensity-modulated radiotherapy and brachytherapy dose matrices by overcoming the challenges imposed by planning systems . For this task, Bice et al  have used image coregistration employing a post-implant CT and ultrasound volume study. In their study, the dosimetric results of the original implant were overlaid on the salvage plan and the total doses to the target and the OARs were evaluated. The main dosimetric parameter reported was V80, which improved from an average of 56% to 93%.
We suggest that LDR seed reimplantation for suboptimal implants presents certain advantages over an EBRT boost. Therefore, when compared with an EBRT boost, LDR top-up offers a better dosimetric outcome owing to more accurate targeting of the cold spots. In addition, the dosimetry and radiobiology of a composite LDR brachytherapy/EBRT plan still remains a challenge for many treatment planning systems. Regarding the dose to OARs, with LDR top-up, the added urethral dose can be kept to a minimum with careful placement of seeds, as compared with EBRT when the urethra would almost certainly be included in the radiation field. Also, an advantage from the patient’s perspective is that only one visit is required for treatment, instead of several visits, which a fractionated EBRT would require.
A salvage reimplantation technique for suboptimal implants was reported by Hughes et al  using fluoroscopic guidance, whereby an underdosed base owing to systematic seed misplacement led to a V100=46%. For better localisation of the seeds to be implanted, a grid was superimposed on the original post-implant CT images and digitised into the planning system to facilitate dosimetric evaluation (similar to planning on transrectal ultrasound images). The reimplantation results showed V100=98%, although with increased urinary morbidity.
Keyes et al  have described the salvage implants undertaken on seven patients with poor initial dosimetry. Their planning simulation principle was similar to the one reported in the current work, whereby extra seeds were simulated on the underdosed areas of prostate on post-implant CT images, maintaining the original 5-mm distance between seeds. A composite dose was determined by adding the doses for the two implants together, without taking into account the radioactive decay of the initially implanted seeds.
A recent case report has described the salvage reimplantation of a suboptimally implanted prostate cancer patient by employing post-implant CT guidance and real-time intra-operative planning using loose seeds . PID revealed an underdosed base, leading to D90=92.9 Gy and V100=71%, which after the addition of 20 seeds increased to D90=173.24 Gy and V100=96.88%. The flexibility of intra-operative planning has allowed continuous adjustments of seed positioning, thus making this technique suitable for salvage implants.
As a drawback, LDR top-up is labour intensive and involves a large team of professionals to actively participate in treatment preparation and delivery. Furthermore, the patient requires a second general anaesthetic, which might be a problem for those patients with an underlying condition. Also, since patients might live in other cities, thus a long distance from the hospital, top-up sessions would have to accommodate their travel arrangements, which might sometimes delay the boost implant. Nevertheless, the same problem might be encountered if EBRT is chosen as boost treatment.
PID is a valuable tool in assessing the quality of an implant. Despite all team efforts taken to achieve high-quality implants, every so often implants can be suboptimal owing to technical-, staff experience- or patient-related matters.
To correct the dosimetry of an unsatisfactory implant, a top-up procedure was implemented in our centre, involving the implantation of extra seeds to account for the cold spots and other geographical misses. A description of the methods used to overcome the technical difficulties in planning top-up implants has been presented.
Although the final target dosimetry achieved was satisfactory, the procedure was compromised by higher doses being delivered to the OARs than the recommended limits. The affected patients are kept under close observation to identify and treat any normal tissue toxicities.
In conclusion, we found that iodine-125 seed top-up is a feasible method for improving the quality of a previous suboptimal implant.
The authors would like to acknowledge the radiation oncologists, urologists, radiation therapists, medical physicists, nursing staff and all other members of the prostate brachytherapy team at the Royal Adelaide Hospital for their dedication to the prostate brachytherapy service over the years.
The information comes from:https://www.ncbi.nlm.nih.gov/pmc/articles/PMC3664978/
Abstract: The importance of personalized medicine has been growing, mainly due to a more urgent need to avoid unnecessary and expensive treatments. In nuclear medicine, the theranostic approach is an established tool for specific molecular targeting, both for diagnostics and therapy. The visualization of potential targets can help predict if a patient will benefit from a particular treatment. Thanks to the quick development of radiopharmaceuticals and diagnostic techniques, the use of theranostic agents has been continually increasing. In this article, important milestones of nuclear therapies and diagnostics in the context of theranostics are highlighted. It begins with a well-known radioiodine therapy in patients with thyroid cancer and then progresses through various approaches for the treatment of advanced cancer with targeted therapies. The aim of this review was to provide a summary of background knowledge and current applications, and to identify the advantages of targeted therapies and imaging in nuclear medicine practices. Keywords:theranostics, nuclear medicine, personalized medicine, PET/CT, therapy, diagnostics
Introduction: goals of theranostics in nuclear medicine
Challenges in modern oncology include the fact that patients are getting older and are typically unfit for conventional chemotherapy regimens because of comorbidities or poor performance status.1 Furthermore, the occurrence of side effects may aggravate treatment compliance in both young and elderly patients.2 To manage these problems, it is important to improve patient selection, reduce side effects, and enhance therapeutic efficacy. Taking these factors into consideration, the combination of targeted cancer imaging and therapy is a considerable achievement for personalized medicine.
The theranostic approach in nuclear medicine couples diagnostic imaging and therapy using the same molecule or at least very similar molecules (Figure 1), which are either radiolabeled differently or given in different dosages. For example, iodine-131 and lutetium-177 are gamma and beta emitters; thus, these agents can be used for both imaging and therapy. Furthermore, different isotopes of the same element, for example, iodine-123 (gamma emitter) and iodine-131 (gamma and beta emitters), can also be used for theranostic purposes.3,4 Newer examples are yttrium-86/yttrium-90 or terbium isotopes (Tb): 152Tb (beta plus emitter), 155Tb (gamma emitter), 149Tb (alpha emitter), and 161Tb (beta minus particle).5,6
Figure 1 The theranostic principle in nuclear medicine involves combining diagnostic imaging and therapy with the same molecule, which is radiolabeled differently, or administered in other dosages. In case of radioiodine therapy (RAI), the radioisotope (131I or 123I) can be directly mediated by the sodium-iodide symporter in the thyroid cells. In other cases, it can be more complex. The image shows a simplified model of a radiopharmaceutical, which consists of a binding molecule that binds the target, and a linking molecule, which binds the radioisotope. Examples of such theranostic molecules are DOTA-TOC, DOTA-TATE, and PSMA-617.
The detection of potential targets can help predict whether a patient will benefit from a particular treatment. Theranostics can be useful for estimating the potential response and eventual toxicity. During the treatment, theranostics can be applied in monitoring the therapy course. However, one cause of concern is the safety of high cumulative doses of radioactive agents after multiple repeated cycles. For instance, reports concerning irreversible high-grade toxicity followed soon after the first treatments with radioiodine (iodine-131).7 However, there have been remarkable advances in nuclear medicine, especially in the field of targeted therapies. After a proper preselection of candidates, targeted nuclear therapies have proven to be effective in the majority of cases with a favorable safety profile.8–11 Recent studies have shown no evidence of grade 3/4 toxicity in patients with neuroendocrine tumors after repeated (≥8 cycles) radiopeptide therapy with a cumulative dose of up to 97 GBq.12 Good tolerability has also been observed in patients with prostate cancer (PC) after a cumulative dose of 36 GBq and up to six cycles of radioligand therapy.13,14
Nuclear imaging utilizes gamma and positron emitters (β+). Gamma emitters, such as technetium-99m (99mTc) or iodine-123 (123I), can be located using gamma cameras (planar imaging) or SPECT (single photon emission computed tomography).15 However, better resolution can be achieved via PET (positron emission tomography) using positron emitters, such as gallium-68 (68Ga) and fluor-18 (18F).16
Most therapeutic radiopharmaceuticals are labeled with beta-emitting isotopes (β−). The tissue penetration of these particles is proportional to the energy of the radioisotopes.17 Beta particles have a potential cytocidal effect, but they also spare the surrounding healthy tissue due to having a tissue penetration of only a few millimeters.8 Commonly used beta emitters in routine nuclear oncology practices include lutetium-177 (177Lu, tissue penetration: 0.5–0.6 mm, maximum: 2 mm, 497 keV, half-life: 6.7 days) and yttrium-90 (90Y, tissue penetration: mean 2.5 mm, maximum: 11 mm, 935 keV, half-life: 64 hours).8,10,13,18–20
The first theranostic radiopharmaceutical in nuclear medicine history was radioiodine, which was used for therapy and imaging in thyroid diseases.21 Since then, the use of theranostic agents has been consistently increasing. Nuclear targeted therapies play an essential role, especially in patients with advanced neuroendocrine tumors, such as gastroenteropancreatic (GEP) tumors, bronchopulmonary neuroendocrine tumors, pheochromocytoma, and neuroblastoma.10,11,20,22–27 Furthermore, there are positive results with radioligand therapies in metastatic PC and metastatic melanoma.8,13,19,28,29
The aim of this review was to discuss the most important milestones of nuclear theranostics in current practice (Table 1), and to provide a summarized background and overview of current applications and advantages of targeted therapies and imaging.
Table 1 Overview of theranostic agentsAbbreviations: mIBG, metaiodobenzylguanidine; SSA, somatostatin analogs; SSTR, somatostatin receptors; NEN, neuroendocrine neoplasia; GEP, gastroenteropancreatic system; SPECT, single photon emission computed tomography; PET, positron emission tomography.
Radioiodine therapy: “the gold standard” in thyroid diseases
Iodine (stable isotope: iodine-127) is taken up by the thyroid gland for the production of thyroid hormones, namely, thyroxine (T4) and triiodothyronine (T3).30 Thyroid hormones are vital for the embryonic and neonatal development of the brain, normal growth, and metabolic balance.9,31–33 In 1936, Dr Saul Hertz, director of the Thyroid Clinic in Massachusetts (1931–1943), developed the idea of administering radioactive iodine in patients with thyroid diseases. Iodine and external beam radiation were well-known tools in thyroid disease therapy, but the combination of these was a considerably innovative approach. This idea followed a few years of preclinical studies in collaboration with the Massachusetts Institute of Technology (MIT), where the first cyclotron for medical use had been built. The MIT Cyclotron produced 90% iodine-130 (130I, half-life: 12 hours) and 10% iodine-131 (131I, half time: 8 days). Subsequently, on March 31, 1941, Dr Hertz treated the first patient with radioiodine (130I).21,34,35
The first radioiodine therapy (RAI) with 131I in patients with thyroid cancer (TC) was undertaken by Seidlin et al in 1946.36 This group studied the use of RAI in patients with metastatic thyroid carcinomas.36,37 Seidlin et al also reported one of the first cases of acute myeloid leukemia after repeated RAI treatments.7
To date, 131I is still the gold standard for the therapy and diagnosis of differentiated TC.38 It is a low-cost nuclear reactor product from the neutron bombardment of tellurium-131. 131I combines the characteristics of a beta (β−, approximately 90% of the radiation, mean: 192 keV, mean tissue penetration: 0.4 mm) and gamma (approximately 10% of the radiation, mean: 383 keV) emitter. In this way, it irradiates the TC and the thyroid remnant from the inside and, at the same time, targeted lesions can be visualized using a gamma camera or SPECT.9,32,33 Another radioisotope is 123I, which is a pure gamma emitter and is used for pre- and post-therapeutic diagnostics. The advantages of imaging with 123I include a higher quality of whole-body scans, which improves the sensitivity in detecting thyroid remnants.3,4
Figure 2A shows the initial 131I-planar images of a 74-year-old female with metastatic TC (lung, bone, and intracranial soft-tissue metastases). The level of tumor marker thyroglobulin (Tg) before RAI was 572 ng/mL. After two administrations of RAI (cumulative activity: 14.3 GBq), the patient was in complete remission with a Tg level of 0.2 ng/mL (Figure 2B).
Figure 2 (A) Initial 131I-planar images of a 74-year-old female with metastatic thyroid cancer (lung, bone, and intracranial soft-tissue metastases, marked with arrows). The tumor marker thyroglobulin (Tg) before radioiodine therapy (RAI) was 572 ng/mL. (B) After two administrations of RAI (cumulative activity: 14.3 GBq), the patient was in complete remission with the Tg at 0.2 ng/mL. The planar images show only a physiological uptake of the radiotracer in the gastrointestinal tract and pharyngeal mucosa (marked with an asterisk).
Diagnostics and therapy with metaiodobenzylguanidine
Metaiodobenzylguanidine (mIBG), or iobenguane, is a molecule similar to noradrenaline and enters neuroendocrine cells from the sympathetic nervous system, either by endocytosis or passive diffusion before being stored in neurosecretory granules.39 Among the used radiolabeled molecules, [123I]I-mIBG has a lower gamma energy than [131I]I-mIBG (159 keV vs 360 keV), which makes it more suitable for planar imaging/SPECT. Furthermore, the pure gamma emitter [123I]I-mIBG consists of a shorter half-time of 13 hours compared to 8 days for the combined beta and gamma emitter [131I]I-mIBG, leading to a smaller radiation burden. Thus, higher quantities of [123I]I-mIBG can be injected.40 Both [131I]I-mIBG and [123I]I-mIBG are used in mIBG scintigraphy for the detection of neuroendocrine tumors, such as neuroblastomas, pheochromocytomas, paragangliomas, medullary thyroid carcinomas, and other neuroendocrine neoplasias.41 In patients with inoperable or advanced tumors with distant metastases, mIBG imaging plays an essential role in response assessment after therapy and in the evaluation of potential [131I]I-mIBG therapy.42,43 In patients with neuroblastoma and pheochromocytoma, [123I]I-mIBG has high sensitivity (97% and 94%) and specificity (up to 96% and 92%), respectively.44–46 If available, [124I]I-mIBG-PET can be equally used for the planning of mIBG targeted therapy.47,48
Targeted therapy with [131I]I-mIBG presents an encouraging efficacy with tolerable toxicity in relapsed or refractory neuroblastomas with response rates between 20% and 40% being used alone or in combination with high-dose chemotherapy followed by autologous stem cell transplantation.23–25,49 Recently, the NB2004 Trial for Risk Adapted Treatment of Children with Neuroblastoma closed, and further analysis of the usage of mIBG therapy as a first-line therapy is outstanding.23,25
Palliation of bone metastases
Radiolabeled phosphonates have a high bone affinity and can be used for imaging and palliation of painful bone metastases. Depending on the degree of osseous metabolism, the tracer accumulates via adhesion to bones and, preferably, to osteoblastic bone metastases.8
Therapy planning requires a bone scintigraphy with technetium-99m-hydroxyethylidene diphosphonate (HEDP) to estimate the metabolism and the extent of the metastases involvement. Bisphosphonate HEDP can be labeled for therapy either with rhenium-186 (beta-emitter, half-life: 89 hours, 1.1 MeV maximal energy, maximal range: 4.6 mm) or rhenium-188 (beta-emitter [to 85%, 2.1 MeV] and gamma-emitter [to 15%,155 keV], half-life: 16.8 hours, maximal range in soft tissue: 10 mm).50
Both agents induce pain relief in ≥90% of patients.51 However, rhenium-188 is of particular interest because it can be cost-effectively produced using a tungsten-188/rhenium-188 generator. The gamma component of the agent allows post-therapeutic imaging of good quality.8 Furthermore, patients with prostatic cancer have shown, after two [188Re]Re-HEDP injections, a significantly higher response rate (92%), a prostate-specific antigen (PSA) decline >50% (38%), and a longer PFS (7 months) compared to patients who have had a single injection.52,53 Biersack et al showed an improved OS of 15.7 months in patients after repeated injections (≥3 injections) compared to patients with a single therapy.54 Besides transient bone marrow toxicity (thrombocytopenia grade III), the reported toxicity is low to moderate.51
New promising radiopharmaceuticals for bone palliation therapy include radiolabeled complexes of zoledronic acid. Zoledronic acid belongs to a new, most potent generation of bisphosphonates with cyclic side chains. The bone affinity of zoledronic acid labeled with scandium-46 or lutetium-177 has shown excellent absorption (98% for [177Lu]Lu-zoledronate and 82% for [46Sc]Sc-zoledronate), which is much higher than of bisphosphonates labeled with samarium-153 (maximum: 67%).55
Radiolabeled somatostatin analogs
Neuroendocrine neoplasia (NEN) of the GEP system originates most frequently from the pancreas, jejunum, ileum, cecum, rectum, appendix, and colon. The common characteristic of all GEP-NEN is the compound features of endocrine and nerve cells.56–58 Well-differentiated NEN overexpresses somatostatin receptors (SSTRs), especially the SSTR-2 subtype. Thus, SSTRs are theranostic targets in NEN that have been known for almost three decades and have become well established.59–61
SSTR imaging is necessary for staging, therapy planning, and follow-up. In the PET diagnostics, there are three routine somatostatin analog (SSA) tracers labelled with gallium-68: DOTA-TATE, DOTA-TOC, and DOTA-NOC. All three tracers bind specifically with SSTRs.62,63 Gallium-68-labeled SSTRs have high sensitivity (82%–97%) and specificity (80%–92%) in the detection of small primary tumors or metastases of GEP-NEN.64–66 Alternatively, a conventional whole-body scintigraphy (planat plus SPECT) or SPECT alone can be performed using indium-111-labeled SSTR, which is available at many institutions. Even if the 111In imaging does not have as good a quality as PET, this method has been shown to be superior to chromogranin A measurements alone for the management of GEP-NET patients and is better than computed tomography or ultrasound.67,68
Peptide receptor radionuclide therapy (PRRT) is a systemic therapy in patients with advanced metastatic NEN. PRRT requires a good tumor uptake in the SSTR imaging. For therapeutic purposes, the peptides DOTA-TATE and DOTA-TOC can be labeled with either 90Y or 177Lu. Because of the renal excretion and the proximal tubular reabsorption of these tracers, the kidneys are one of the primary limiting organs for this therapy.69 Due to the smaller range (2 mm) and lower energy, 177Lu is less nephrotoxic than 90Y (range maximum: 11 mm, pure beta emitter, higher energy). 177Lu is also less hematotoxic.70 This allows for the performance of several PRRT cycles without any relevant toxicity. Furthermore, the overall survival is not significantly different between the therapies. For that reason, [177Lu]Lu-DOTA-TATE and DOTA-TOC are, in many centers, the preferred agents for NEN therapy.10,11,71,72
PRRT has been increasingly gaining attention. The first randomized controlled phase III study, NETTER-1 (started 2012), compared the standard therapy of Sandostatin® LAR (somatostatin analog) with PRRT in patients with midgut NEN. The study showed a significant clinical benefit from the therapy, achieving a prolonged progression-free survival (median not reached, approximately 40 months, p<0.001), an overall response of 18%, and a presumably longer overall survival (median not reached, p=0.004).11
Figure 3 represents the image of a male patient with a neuroendocrine tumor (unknown origin), with a recurrence of the disease nearly 2 years after the baseline PRRT. After a salvage therapy of PRRT, the patient had a partial response.
Figure 3 (A) [68Ga]Ga-DOTA-TOC-image* of a male patient with a neuroendocrine tumor (unknown origin) 23 months after the baseline PRRT (three cycles, cumulative activity: 19.6 GBq). The patient had a recurrence of the disease with multiple metastases in the bone, lung, liver, and lymph nodes (marked with arrows). (B) After another course of PRRT (three cycles, cumulative: 43.4 GBq), the PET images showed a partial response.Note: *Maximum-intensity projection (MIP) PET image is a visualization technique that provides an initial overview of the case.Abbreviations: PET, positron emission tomography; PRRT, peptide receptor radionuclide therapy.
PC is the most common cancer in men in Western countries.73 Hormone- and chemotherapy-refractory patients with metastatic PC have a poor prognosis.8,74,75 The main cause of death in these patients is progression to the androgen-independent stage.8
PC cells overexpress prostate-specific membrane antigen (PSMA) on the cell surface.76–79 There are several available radiopharmaceuticals that target PSMA including [68Ga]Ga-PSMA-HBED-CC (also known as [68Ga]Ga-PSMA-11 [PET]), a monoclonal antibody (mAb) [177Lu]Lu/[90Y]Y-J591 (therapy), [123I]I-MIP-1072 (planar/SPECT), [131I]I-MIP-1095 (therapy), and the theranostic agents PSMA-I&T and DKFZ-PSMA-617 (PSMA-617), which are labeled with 68Ga for PET or with 177Lu for therapy.
The specificity of the two commercially available PET-tracers, [68Ga]Ga-PSMA-617 and [68Ga]Ga-PSMA-11, is similar – 99% and 100%, respectively.80–82 However, due to slower kinetics, PSMA-617 was suggested for labeling with the long-half-life Lutetium-177 for therapy and, therefore, there is sparse systematic evaluation of [68Ga]Ga-PSMA-617 for diagnostics.83 The high kidney uptake of PSMA-11 in the kidneys makes it unsuitable for therapy and does not fit the definition “theranostics.”84 Still, it is at issue if molecules with affinity to identical target structures might already present a “theranostic surrogate” even if more than the isotope differs between the diagnostics and the therapeutics.
The latest studies have shown that treatment with [177Lu]Lu-PSMA-617 is effective and well tolerated. In fact, nearly 70% of patients have benefited from this therapy.13,28,85–90 Dosimetry studies have shown that the most critical organs are the kidneys, with a maximum kidney radiation dose of 0.88 Gy/GBq for [177Lu]Lu-PSMA-617 and 0.93 Gy/GBq for [177Lu]Lu-PSMA-I&T.91,92 Okamoto et al stated that a cumulative activity of 40 GBq [177Lu]Lu-PSMA-I&T could be safely applied in patients.91 Furthermore, [177Lu]Lu-PSMA-617 has been shown to produce no relevant increase in renal toxicity in the salvage setting or in patients with kidney radiation doses >19 Gy.14
Figure 4A shows the pre-therapeutic [68Ga]Ga-PSMA-11 images of a patient with multiple lymph node, peritoneal, and bone metastases (arrows), and a history of chemotherapy (first and second line), enzalutamide, and abiraterone. After three cycles of [177Lu]Lu-PSMA-617, the follow-up images show a very good response with a substantial PSA decline (Figure 4B).
Figure 4 (A) Pre-therapeutic [68Ga]Ga-PSMA-11 images* of a patient with multiple lymph node, peritoneal, and bone metastases (arrows), and history of chemotherapy (first and second line) with enzalutamide and abiraterone. (B) After three cycles of [177Lu]Lu-PSMA-617, the follow-up images showed a very good response with a substantial PSA decline.Note: *Maximum-intensity projection (MIP) PET image is a visualization technique that provides an initial overview of the case.Abbreviations: PET, positron emission tomography; PSA, prostate-specific antigen.
Melanin targeting in patients with metastatic melanoma
A promising approach in patients with metastatic melanoma is the specific targeting of melanin. The newly developed theranostic agents include [123I]I-/[131I]I-BA52 and [18F]F-/[131I]F-ICF15002, which may play a considerable role in the future.
BA52 is a melanin-binding benzamide. Labeled with 123I, it shows a specific binding of the pigmented metastases in planar imaging/SPECT and can help select patients who would probably benefit from the therapy. In a pilot study, [131I]I-BA52 was effective in three of five patients who were treated with more than 4.3 GBq.29
ICF15002 is an arylcarboxamide derivative and, as a small molecule, can passively enter the cell and bind to melanin. The PET tracer is radiolabeled with 18F and [131I]I-ICF15002 is designed for the therapy. However, both tracers are still in the preclinical phase of their studies. One major problem may be the absorbed dose in melanin-rich tissues, such as skin, dark eyes, and the brain. For instance, in a murine model, there was a 30% decrease of the retinal thickness after two cycles of [131I]I-ICF15002.93
In nuclear medicine, theranostics combine diagnostic imaging and therapy with the same, but differently radiolabeled, molecule, or the same agent in different dosages. The visualization of potential targets can help predict if a patient would benefit from a particular treatment. In properly preselected patients, targeted nuclear therapies have proven to be effective with a favorable safety profile. To conclude, the combination of targeted cancer imaging and therapy is a considerable contribution to personalized medicine and may play an increasingly important role in the future.
The information comes from:https://www.dovepress.com/theranostics-in-nuclear-medicine-practice-peer-reviewed-fulltext-article-OTT
Background: This review provides a comprehensive summary of the production of 177Lu to meet expected future research and clinical demands. Availability of options represents the cornerstone for sustainable growth for the routine production of adequate activity levels of 177Lu having the required quality for preparation of a variety of 177Lu-labeled radiopharmaceuticals. The tremendous prospects associated with production of 177Lu for use in targeted radionuclide therapy (TRT) dictate that a holistic consideration should evaluate all governing factors that determine its success. Methods: While both “direct” and “indirect” reactor production routes offer the possibility for sustainable 177Lu availability, there are several issues and challenges that must be considered to realize the full potential of these production strategies. Results: This article presents a mini review on the latest developments, current status, key challenges and possibilities for the near future. Conclusion: A broad understanding and discussion of the issues associated with 177Lu production and processing approaches would not only ensure sustained growth and future expansion for the availability and use of 177Lu-labeled radiopharmaceuticals, but also help future developments.
Over the last several years, the 177Lu radionuclide has attracted considerable attention and exhibited great promise in the research, commercial and clinical communities for use in a variety of therapeutic procedures [1–8]. Despite being a late entrant, 177Lu has not only consolidated its potential, but also established a strong foothold at the forefront of TRT. In a relatively short time span, 177Lu has virtually pervaded all areas of in vivo radionuclide therapy and may be poised to become a key therapeutic radionuclide of choice for TRT. The growing interest in the use of 177Lu in targeted molecular therapies has primarily developed from recent unprecedented advances in molecular and cell biology, which include the use of peptides targeted to cell surface receptors, which are overexpressed on the surface of tumor cells.
Lutetium-177 decays in 76 % of events (E β(max) =0.497 MeV) to the stable ground state of 177Hf with a half-life of 6.65 days and decays in 9.7 % of events (E β(max) =0.384 MeV) and 12 % of the time (E β(max) = 0.176 MeV) to an excited state of 177Hf that lies 0.24967 MeV and 0.32132 MeV above the ground state, which de-excites to the ground state with the photon emission. During these radioactive decay events, 177Lu emits β- particles with an E β(max) of 497 keV (78.6 %), 384 keV (9.1 %) and 176 keV (12.2 %) and low-energy gamma photons [Eγ = 113 keV (6.6 %), 208 keV (11 %)] [9, 10]. A simplified decay scheme for 177Lu is shown in Fig. 1.
Simplified decay scheme of 177Lu
Increasing use of 177Lu in nuclear medicine procedures has been impressive, and widespread applications of 177Lu therapeutic agents have not only stimulated progress of TRT, but have also been responsible for stimulating the growth of these therapeutic methods. The utility of 177Lu is thus continually evolving, well entrenched in the arena of targeted radionuclide therapy, and has unveiled a broad spectrum of 177Lu-labeled therapeutic radiopharmaceuticals for treating a wide range of diseases. It is surmised by many that diffusion of 177Lu into nuclear medicine has not only brought spectacular developments in radionuclide therapy, but has also prompted a perceptible shift of the radiotherapeutic method toward the treatment of some diseases. The remarkable prospects and impetus associated with the use of 177Lu-labeled radiopharmaceuticals in TRT have been the major factors evoking excitement among researchers and capturing the imagination of the clinical community thanks to advances in molecular and cellular biology.
While myriad factors contribute to the success of 177Lu-labeled radiopharmaceuticals, the cost-effective availability of sufficient activity levels of 177Lu that have the required quality is a key determinant underpinning the success of using 177Lu in in vivo targeted therapy. The recent surge of interest in the use of 177Lu in TRT has been the motivation to provide this review on the production and processing of this emerging radionuclide. This article focuses on a discussion of the 177Lu production strategies that are currently used and other approaches that may have considerable potential in the foreseeable future.
The striking diffusion and exciting perspectives of 177Lu in TRT are primarily attributed to the following.
The mean penetration range of β− particles emitted by 177Lu in soft tissue is 670 μm, making this radionuclide ideal for delivering energy to small volumes, including micrometastatic disease, and tumor cells near the surface of cavities. Lutetium-177 is found to be effective in localizing cytotoxic radiation in relatively small areas and proficient in destroying small tumors as well as metastatic lesions (typically less than 3 mm diameter) with less damage to surrounding normal tissue.
The emission of low-energy gamma photons enables imaging the biodistribution and excretion kinetics with the same radiolabeled preparation used for therapy and allows dosimetry to be performed before and during treatment as well. This property is important for “personalized” medicine for the development of “theranostic” agents for combined diagnostic and therapeutic use that can deliver therapy to individual cells in affected tissues.
The emission of moderate-energy beta β- particles as well as low-energy gamma photons results in a relatively low radiation dose and therefore offers the potential to handle relatively high 177Lu activity levels during radiopharmaceutical preparation and formulation of radiopharmaceuticals as well as during patient administration.
Lutetium exclusively exists in the +3 oxidation state, which precludes any solution chemistry reduction-oxidation complications and commonly forms nine coordination complexes. This property therefore provides the potential for radiolabeling a variety of molecular carriers, which include small molecules, and peptides, proteins and antibodies with the specific desired characteristics for therapy. The chemical characteristics of Lu+3 are suitable for peptide and protein radiolabeling by attachment of a bifunctional chelating agent (BFCA) through a metabolically resistant covalent bond.
The 6.65-day half-life of 177Lu offers extended time periods, which may be required for the use of more sophisticated procedures to radiolabel and purify 177Lu-labeled radiopharmaceuticals, and for performing quality control and administration. The use of a longer lived therapeutic radionuclide such as 177Lu is particularly well suited for the radiolabeling of antibodies that have slow targeting kinetics.
The relatively long 6.65-day physical half-life of 177Lu not only minimizes decay loss, which may be encountered during the transportation and distribution to users, but also provides excellent logistical advantages for shipment to sites distant from the reactor production facility as well as radionuclide-processing facilities.
A wide range of 177Lu radiopharmaceuticals has been successfully developed and evaluated. The in vivo applications of key 177Lu radiopharmaceuticals for a variety of therapeutic procedures include peptide receptor radionuclide therapy [11–26], bone pain palliation [27–33], radiation synovectomy [34–39] and radioimmonutherapy [40–46]. There is a steadily expanding list of 177Lu-labeled radiopharmaceuticals that is currently being evaluated at the preclinical research or at product development stages; these may potentially be used in vivo in humans for evaluation for radionuclide therapy [1–3]. A summary of key 177Lu-labeled radiopharmaceuticals currently used in TRT is depicted in Table 1.
Key examples of current 177Lu-labeled radiopharmaceuticals
The opportunity for production of 177Lu in research reactors throughout the world indicates that all pertinent factors should be evaluated and assessed. Essentially every conceivable production and processing strategy has been exploited with a view to obtain 177Lu in a chemical form having acceptable radionuclidic and radiochemical purity. Since the inherent success of any production strategy requires a thorough knowledge of all the pertinent key factors, the issues underlying the utility of 177Lu in TRT are discussed below.
Before discussing production of 177Lu in detail, it is relevant to discuss the importance of the specific activity (SA) of 177Lu, since this key factor dictates its utility for TRT.
Radionuclides have maximum theoretical specific activity values referred to as “carrier-free” when all the atoms contain one isotope of the element. Carrier free (CF) thus denotes a radionuclide having 100 % isotopic abundance, i.e., free from any stable isotopes. A radionuclide is characterized as no carrier added (NCA or n.c.a.) to which no carrier atoms have been added and for which precautions have been taken to minimize contamination with stable isotopes of the element in question. It does not necessarily mean, however, 100 % isotopic abundance. ‘Carrier free’ is an idealistic situation and hence the current term used is ‘no carrier added (NCA)’ for a preparation having a specific activity value that approaches the calculated maximum theoretical specific activity.
The theoretical specific activity of ‘carrier-free (CF)’ 177Lu is calculated from the following equation:
N →, number of 177Lu atoms.
λ →, decay constant of 177Lu.
T½ →, half life of 177Lu.
Neutron-capture characteristics, target impurities, secondary nuclear reactions, target “burn-up” and post-irradiation processing/cooling periods are the main parameters affecting the SA of the 177Lu product.
Conscientious harnessing of the nuclear and chemical characteristics of 177Lu in conjunction with the advancement in molecular and cellular biology has not only stimulated the progress of radionuclide therapy, but has also driven the field. As discussed earlier, the utility of 177Lu in radionuclide therapy has undergone rapid and continual evolutionary cycles. Lutetium-177-labeled therapeutic radiopharmaceuticals comprising small molecules, large biomolecules and particles are currently being evaluated for myriad of clinical applications [11–46].
While targeted radionuclide therapy is primarily based on selecting appropriate radiopharmaceuticals and targeting mechanisms, the number of target sites (receptors, cells, etc.) available for radiopharmaceutical targeting dictates the SA of 177Lu that is appropriate for a particular application. For instance, targeting to trabecular bone is considered a large-capacity site and does not require 177Lu with highly specific activity. For this reason, the 177Lu SA is basically of minimal consequence in applications involving preparation of 177Lu-labeled radiopharmaceuticals used for treatment of bone pain palliation, hepatocellular carcinoma (HCC) and synovectomy, since high masses of the low SA agents are administered. Under this premise, medium-low SA 177Lu produced by the “direct” route described below can generally be used. However, high SA 177Lu is required for other applications involving low capacity sites, which are present in low numbers, such as receptor sites for peptide and antibody therapy. In the field of peptide receptor radionuclide therapy (PRRT) with radiolabeled peptide analogs, use of high SA 177Lu constitutes a necessity owing to the limited concentrations of the different cellular cognate receptors expressed on the tumor cell surface. Similar is the case with radioimmunotherapy (RIT), where the use of radiolabeled monoclonal antibodies targets different tumor-associated antigens and exceeding the mass threshold for pharmacologic activity of the antibody could cause unwanted reactions. For this application, the tumor cell antigen concentrations are overexpressed compared to limited concentrations associated with normal cells. Unlike peptides, monoclonal antibodies are macromolecules with molecular weights of about 150,000 Da, so the specific activity values that can be achieved by adding one or two 177Lu atoms to each molecule will also be low. Hence, RIT also requires high-specific-activity radionuclide preparations.
Both the “direct” and “indirect” reactor production routes can be followed to obtain 177Lu for nuclear medicine applications: The direct production route is based on neutron irradiation of 176Lu targets by the 176Lu(n,γ)177Lu reaction. The indirect 176Yb(n,γ)177Yb → 177Lu production route necessitates a chemical separation of 177Lu from the target 176Yb target atoms.
The direct and indirect reactor routes for production of 177Lu are shown in Fig. 2. Each route has specific advantages and disadvantages, which are elaborated in the following sections. The neutron activation products of natural lutetium and ytterbium targets along with the nuclear decay characteristics of the product radionuclides are given in Table 2.
Two different routes for reactor production of 177Lu
Neutron activation products of natural lutetium and ytterbium targets along with the nuclear decay characteristics of the product radionuclides
The direct production route offers the following advantages.
The least intricate approach to target irradiation in a reactor and requires minor design changes in reactor irradiation and processing facilities.
Offers the potential to use the 176Lu2O3 target, which remains stable under irradiation conditions and is compatible with reactor irradiation.
Irradiated target processing is easy, fast and technically less demanding as simple target dissolution in dilute mineral acid on gentle warming suffices. The facility required for target processing is straightforward to install and maintain.
Has the flexibility to scale the increase or decrease levels of production in response to requirements by adjusting the target size.
Processing generates negligible levels of radioactive waste.
This production method represents the most inexpensive option to obtain 177Lu of requisite purity.
Unlike other medically useful radionuclides, the direct (n,γ) production route often offers the prospect of producing 177Lu with SA adequate for preparing receptor-specific therapeutic radiopharmaceuticals. This is possible because 176Lu has a very high thermal neutron capture cross section (σ = 2090 b, I0 = 1087 b) for formation of 177Lu. The neutron capture cross section of 176Lu does not follow the 1/v law, and there is a strong resonance very close to the thermal region [50, 54].
In spite of the advantages of the direct production route, some concerns that have been raised on the use of this production route include:
With a view to augmenting the 177Lu production yield as well as specific activity, the possibility of using enriched 176Lu targets is necessary owing to the limited natural abundance (2.6 %) of 176Lu in the unenriched target.
The specific activity of 177Lu obtained by this method is generally 740–1,110 GBq (20–30 Ci)/mg versus the theoretical SA value of 4.07 TBq (110 Ci)/mg. This indicates that only 25 % of the atoms are 177Lu, and 75 % consisting of the product mixture are nonradioactive contaminating 175/176Lu atoms. Thus, the maximum obtainable SA that can be achieved only with high-flux reactors is about 70 % of the theoretical value.
These SA values are adequate for preparation of the 177Lu-labled agents used for bone pain palliation, synovectomy, treatment of liver cancer and some other applications.
However, while the directly produced 177Lu 740–1,110 GBq (20–30 Ci)/mg can be sufficient for PRRT, the SA of course decreases with time; therefore, the shelf life of 177Lu obtained by this method is limited for PRRT and for use in other applications that may require higher SA.
A unique feature of this method is the co-production of 177mLu, the presence of which can be associated with radiation protection and waste disposal challenges in some countries, which are discussed in more detail later in this section.
175Lu (97.4 %) and 176Lu (2.6 %) are the two naturally occurring isotopes of lutetium. While only 175Lu is truly “stable,” 176Lu decays by beta decay with a half-life of 4 × 1010 years. While undertaking the production of 177Lu, the Lu2O3 is the preferred chemical form because of its chemical and thermal stability during irradiation and its solubility in dilute mineral acid. There appears to be great interest in the use of an enriched 176Lu target in view of the explicit need to obtain high-specific-activity 177Lu amenable for radionuclide therapy. Additionally, the targets used for production should be of exceptionally high purity as isotopic impurities are likely to decrease the specific radioactivity of the produced 177Lu owing to high target nuclide burn-up during high neutron-flux irradiation.
The following traditional equation is used to project the irradiation yields:
NL176u → number of target atoms 176Lu,
σL176u → neutron capture cross section of 176Lu (cm2),
N177Lu → number of radioactive atoms 177Lu formed
λL177u → decay constant of 177Lu (s−1),
φ → neutron flux (cm−2 s−1),
Integrating Eq. (2) leads to
Owing to the abnormally high neutron absorption cross section (σ = 2065 barn), target burn-up and consumption of product atoms must be considered when undertaking long-term irradiations.
N0L176u→ Initial number of target atoms 176Lu
Equation (4) is generally used, provided the neutron capture cross section in the thermal neutron area is inversely proportional to the neutron velocity (the 1/Vn law). However, owing to the resonance in the cross section of 176Lu (Table 2) in the thermal neutron energy range , the neutron capture cross section of 176Lu does not follow the 1/v law. Under this premise, using the simplified Westcott convention is an effective proposition for the calculation of the reaction rate of 176Lu(n,γ)177Lu in which an additional correction of the activation rate as a function of the thermal neutron flux temperature is necessary to account for the non-abeyance of the 1/Vn law. This is expressed by the factor k [48–50]. Following the simplified Westcott convention and taking into the account the target burn-up of 176Lu as well as 177Lu atoms during irradiation, the accumulation of 177Lu can therefore be expressed by a differential equation:
The solution of Eq. (5) gives rise to
Gth and Gr are, respectively, the thermal and epithermal neutron self-shielding factors (both can be set equal to 1 if diluted samples are irradiated), g(Tn) is the Westcott factor  , r(α)TnT0−−√ is the spectral index, S0(α) = So(Er)− α, where S0 = 1.67 and Er =0.158 eV are constants of 176Lu , and α is a measure of the epithermal flux deviation from the ideal 1/E distribution, where E is the neutron energy.
The activity produced ‘A’ (in Bq) at the end of irradiation can therefore be calculated using the formula:
The ‘k’ value is the so-called ‘k-factor,’ and the value of ‘k’ is reported to be between 1.5–2.5 . The expression of the yield of 177Lu (Eq. 7) is based on the assumption that the neutron flux is highly thermalized and there is practically no contribution of epithermal neutrons toward 177Lu formation. Based on Eq. 7, the variation of yield of 177Lu per mg of target irradiated with the duration of irradiation is shown in Fig. 3 when an enriched (82 % in 176Lu) target is irradiated at a thermal neutron flux of 1.2 × 1014 n.cm-2.s-1. Figure 3 shows that the yield of 177Lu passes through a maximum and then decreases as the time of irradiation increases. With increasing the ‘k’ value, the yield of 177Lu during neutron irradiation increases.
Variation of the specific activity of 177Lu, theoretically calculated using k = 1.5, 2.0 and 2.5, with the duration of irradiation when the enriched (82 % in 176Lu) target is irradiated at a thermal neutron flux of 1.2 × 1014 n.cm-2.s-1r
The time of irradiation at which the maximum activity of 177Lu is achieved is expressed as 
It is evident from the equation  that by increasing the neutron flux and/or k-factor, a shorter period of irradiation is required to achieve the optimum yield of 177Lu. It is worthwhile to point out that an irradiation period equal to tmax does not provide the maximum specific activity owing to the transformation of the target material in the nuclear reaction. As reported by Zhernosekov et al.  , the actual specific activity of 177Lu is different from the value obtained by dividing the production of the yield of 177Lu by the mass of the target irradiated, since the actual mass of lutetium present in the system post irradiation is different from the initial mass of the target irradiated owing to target burn-up. During irradiation, 177Lu absorbs neutron and leads to the formation of 177/178Hf, which results in the accumulation of the 177/178Hf carrier in the target system. While the presence of 177/178Hf(IV) has no consequences in the efficiency of 177Lu(III)-labeling reactions , accumulation of hafnium atoms decreases the specific activity.
Using the burn-up correction, the actual specific activity S (Bq/mol) of 177Lu can be expressed as:
Figure 4 compares the variation of the burn-up corrected specific activity of 177Lu calculated using Eq. 9 with these calculated values without taking target burn-up into consideration. It is evident from the estimates that the period of irradiation at which the maximum yield of 177Lu is achieved does not provide the highest available specific activity. Theoretical calculation shows that the available specific activity (burn-up corrected) of 177Lu passes through a maxima at ~21 days of irradiation when the enriched Lu target (82 % 176Lu) is irradiated at a thermal neutron flux of 1.2 × 1014 n.cm-2.s-1. This duration is significantly higher than the theoretically calculated ‘tmax’ of 177Lu yield, which is ~14 days at the same irradiation condition. Irradiation longer than the ‘tmax’ leads to some loss of activity, but also to an increased 177Lu/176Lu ratio and hence increased specific activity due to burn-up of 176Lu. This theoretical analysis justifies the 21-day irradiation cycle used for 177Lu production in the Dhruva reactor in India [54, 55]. The Indian experience has demonstrated that the theoretically calculated value of the actual or available specific activity of 177Lu after 21 days of continuous irradiation of an enriched Lu target (82 % 176Lu) at a thermal neutron flux of 1.2 × 1014 n.cm-2.s-1 (1,142 GBq/mg, using k = 2.5) is close to the practically obtained value (1,108 ± 24 GBq/mg) [54, 55] .
Comparison of variation of calculated 177Lu specific activity taking into account the burn-up correction with that calculated without burn-up correction (both considering k =2)
While the calculation of 177Lu yield based on the simplified Westcott convention is precise enough, the utility of this computation requires accurate knowledge of the neutron flux parameters of the reactor. The accuracy of the 177Lu irradiation yield calculation strongly depends on the stability of the neutron flux parameters during the target irradiation period. The relatively small variations in the calculated and actual 177Lu-specific activities are mostly due to the variation of the neutron flux levels due to the power level of the reactor operation. It is not practicable to normalize the neutron flux level in the multipurpose research reactor.
Therefore, the maximum obtainable specific activity that could be achieved through the direct production route is about 70 % of the theoretical value; this is only possible for irradiations conducted in high neutron-flux nuclear reactors, which are available in a limited number of countries. It has been reported that it is possible to achieve specific activities of 1,850−2,405 GBq/mg (50−65 Ci/mg) by irradiation in higher flux reactors such as the HIFR reactor at Oak Ridge National Laboratory [56, 57]. Lutetium-177 with specific activity values of >740−1,110 GBq (20–30 Ci)/mg could be produced using an enriched 176Lu target up to approximately 60–80 % in medium flux reactors [54, 55] . The SA values are adequate for all established applications of 177Lu for radionuclide therapy.
While using lutetium oxide enriched in 176Lu up to approximately 60–80 % constitutes a successful paradigm for producing 177Lu of specific activities >740 GBq (20 Ci)/mg amenable to radionuclide therapy, the coproduction of 177mLu with a half-life of 160.1 days owing to the 176Lu(n,γ)177mLu (σ = 2 barn) nuclear reaction has emerged as one factor that may be an impediment restricting its utility in some countries. The 177mLu content in the final product depends not only on the irradiation time, but also on the time elapsed after the end of the irradiation (EOI). Under this premise, it is pertinent to note that owing to the long half-life and low neutron absorption cross section, the activity levels of 177mLu formed will be low but still be of possible concern. Whereas the 177mLu waste issue must be locally resolved, estimates have clearly shown that the resulting radiation dose increase from the presence of 177mLu is insignificant at clinically significant dose levels, at least for PRRT . Under the optimized production conditions, the reported values for the 177mLu/177Lu ratio vary between 0.01 %–0.02 % at EOB .
The presence of 177mLu may raise the following concerns:
Radiation dose: As hospitals are using their 177Lu for the preparation of radiopharmaceuticals up to 1 week after EOB, the 177mLu/177Lu ratio would likely be doubled. A usual therapeutic dose of 177Lu ranges between 7 and 9 GBq. When the 177mLu/177Lu ratio is 0.02 %, this means that a dose includes approximately 1.4–1.8 MBq 177mLu.
Laboratory waste: During the radiolabeling process and treatment, the loss of radioactivity is typically 2 to 5 % of 177Lu activity, which corresponds to levels of 28–90 kBq 177mLu. In view of the permissible release limit for 177mLu waste (10 Bq/g), all laboratory radioactive waste is required to be collected separately and shipped to a radioactive waste management facility where it is allowed to decay. With a half life of 160.1 days, a considerable amount of time is required to decay 177mLu.
Waste water: A patient excretes approximately 80 % of the administered dose (1.45 MBq 177mLu) after administration of 177Lu-labeled octreotide through the urine. The patient-excreted activity in urine and feces must be stored in waste water where there is a considerable chance of accumulation of 177mLu in the radioactive waste water holding tanks. According to the European radiation safety regulation, the maximum permissible radioactive concentration of 177mLu in the municipal sewage is 50 kBq/m3 . This means that radioactive waste water from the holding tanks needs to be diluted significantly before discharging into the municipal sewage line. The presence of 177mLu might exceed the activity limits alone or with other nuclides (sum activity) in the radioactive waste water holding tanks and must be evaluated in each case.
While the radiation dose to patients from 177mLu (0.01 %–0.02 %) is of little consequence , the problem of safe handling and disposal of the residual quantities of 177mLu by the hospital user may emerge as a major roadblock, which can hardly be circumvented, in principal, through the storage of the radioactive wastes that is customary in hospitals. Despite the above-mentioned drawbacks, however, 177Lu obtained from the (n,γ) route is preferred by many hospitals owing to the cost-effective availability of acceptable quality and quantities on demand. This can be seen as the window of opportunity to lay the basis for realizing the widespread radiopharmeutical use of 177Lu.
The indirect production route offers the following advantages:
The highest >2.96 TBq (80 Ci)/mg vs. theoretical 4.07 TBq(110 Ci)/mg specific activity of 177Lu is attainable by this production route.
Offers the potential to provide 177Lu of the highest possible radionuclide purity.
The presence of long-lived radioactive impurities (e.g., 177mLu, <10–5 %) is precluded (below the detection limit) and therefore associated with minimum radiation protection and waste disposal issues.
Specific activity is independent of neutron flux.
Offers satisfactory radiolabeling performance.
The 177Lu obtained by this method has a longer shelf-life (up to 2 weeks) owing to no appreciable decrease in specific activity.
However, this production route also has the following shortcomings, which may be expected to obstruct the path toward widescale utility.
Low production yields due to the poor 176Yb thermal neutron reaction cross section (2.5 barn) as compared to the 2090 barn for the “direct” production from 176Lu.
The effective separation of micro amounts of 177Lu from macro amounts of the irradiated Yb target is not only challenging, but also requires an elaborate radiochemical separation as well as purification procedure.
Generates significant amounts of radioactive waste.
By far, this method of production emerges as the most expensive option to obtain 177Lu of requisite purity.
Not only requires an enriched 176Yb target but also its recovery and recycling.
Despite the above-mentioned drawbacks, there are tremendous prospects associated with the use of NCA 177Lu in TRT; hence, this production route is being aggressively pursued by several institutions. The inherent success of this production route resides in the development of an effective strategy for the efficient separation of pure 177Lu from bulky masses of the neutron irradiated Yb target since Yb follows an identical coordination chemistry with the chelating agents used for the preparation of Lu-based radiopharmaceuticals as well as successful recovery of the Yb target for recycling.
As described in Table 2, natural ytterbium consists of a mixture of seven stable isotopes, including 168Yb, 170Yb, 171Yb, 172Yb, 173Yb, 174Yb and 176Yb, among which 174Yb is the most abundant.
Using natural Yb is a major deterrent because of the following:
Neutron irradiation of natural Yb will lead to the co-production of 169Yb (T½ = 32.026 d) and 175Yb (T½ =4.185 d). The presence of these contaminants will not only complicate the irradiated target handling owing to augmentation of the radiation dose, but also involves higher shielding requirements. The radiation dose to the chemical reagents used for sequestering 177Lu will be significantly higher and may lead to radiation degradation.
While the cooling of the irradiated target is an effective measure to reduce the contribution of 169Yb and 175Yb, this will reduce the yield of 177Lu owing to radioactive decay.
Decay of 175Yb via β- emission to 175Lu leads to the accumulation of stable lutetium, which will decrease the specific activity of the separated 177Lu.
In view of these considerations, assessing the potential of enriched 176Yb (up to ~97 %) is not only an interesting prospect, but may also be viewed as a necessary one. Owing to the low target burn-up of enriched 176Yb during production, development of a process for the recovery of the unused enriched target is one of the key factors that would be expected to ultimately contribute to the economic success of 177Lu production by this indirect route.
While the use of a metallic target is successful for neutron irradiation, using ytterbium metal as a target is precluded as it readily oxidizes in air and under oxygen. Furthermore, the implicit need to use concentrated acid to dissolve irradiated Yb metal continues to thwart efforts toward its utilization as a target for neutron irradiation. In this context, the use of Yb2O3 is the only practical choice since it not only possesses sufficient chemical and thermal stability under reactor irradiation, but also allows easy post-irradiation processing by simple target dissolution in dilute acid.
In this case, the net production rate of nuclide 177Lu is given by
Assuming that the number of target atoms, NYb, remains constant (no considerable
target burn-up) and NYb = NLu = 0 at t = 0 (start of irradiation), integration of Eq. (1) gives rise to
Here NYb is the initial number of 176Yb atoms, σYb is the cross section of the 176Yb(n,γ)177Yb reaction, ϕ is the neutron flux of the irradiation source, t is the irradiation time, td is the decay time after irradiation, and λYb and λLu are the decay constants of 177Yb and 177Lu, respectively. Figure 5 compares the calculated yields of production of 177Lu by the indirect route at different neutron flux values and different irradiation times.
Calculated yields of 177Lu activity produced via the indirect route at different neutron fluxes and different irradiation times
While the indirect 177Lu production method resides at the interface between many disciplines, the inherent determinant for the success of this production route lies in the selection of an appropriate Yb/Lu radiochemical separation. The isolation and purification of 177Lu from the neutron-irradiated target have been subjects of considerable interest. The technical issues associated with the separation of microscopic levels of 177Lu from the macroscopic levels of the 176Yb target represent a challenging task. With a view to realizing the objective, it is imperative to evaluate the differences in chemical and physical characteristics of the Yb and Lu elements that could be used to obtain 177Lu of the requisite purity and yield.
As shown in Table 3, the chemical properties of Yb and Lu are very similar. As a result of the similar characteristics of the group elements, the stability constants of metal ions with a particular ligand show only slight differences. However, such ligands could provide the potential for the separation of these two ions using either ion-exchange chromatography or a solvent extraction technique. Careful scrutiny of Table 3 reveals that the differences are only observed with the existence of a relatively stable oxidation state +2 for Yb and the high solubility of metallic Yb in mercury.
Chemical and physical characteristics of Lu and Yb
Due to the fully filled 4f subshell, unlike lutetium, the oxidation state Yb+2 is relatively stable. The properties of Yb2+ are very similar to group 2 metal cations such as Ca2+ and Sr2+. Therefore, Yb2+ forms an insoluble sulfate, whereas Lu3+ does not. This property has been exploited for separation of the Yb+3 and Lu+3 ions. The principal shortcoming of this method is that the separation is not clean and requires multiple steps in order to achieve satisfactory decontamination.
Ion exchange separation is of course also possible using cation exchangers and elution by complexing agents. In this case the difference in the stability constant of Yb and Lu with the complexing agents is exploited to realize separation. The order of elution of Yb and Lu depends on the values of stability constants of the formed complexes, and the one that forms a strong complex is initially eluted.
In addition, the selective reduction of Yb can be judiciously exploited to achieve separation from Lu. Alkali metal amalgam, which is one of the strongest reducing agents, can be used for the reduction of Yb3+ to Yb2+ as well as Yb2+ to the free element, which can then enter the mercury phase owing to its ability to form amalgam with Hg.
Technological realization of Yb/Lu separation strategies poses several challenges and requires a thorough assessment to evaluate their prospects. The following are some key features that must be taken in to consideration when selecting a separation process.
The chemical separation process chosen must be effective for the separation of micro amounts of 177Lu from macro amounts of Yb.
Separation must be performed rapidly not only to minimize decay losses of 177Lu, but also to reduce the radiation dose to reagents to circumvent radiation degradation.
The separation method selected should be proficient to provide 177Lu with the highest possible decontamination factor from ytterbium.
The process must ensure consistency, reproducibility and high product yield (ideally > 85 %) on a continual basis.
The separation process should be capable of providing 177Lu in a suitable chemical form (ionic form) amenable to radiolabeling with a broad class of carrier molecules.
To undertake production of 177Lu on a weekly basis annually makes it essential to have a very high degree of robustness of the separation process. Nevertheless, the process should be simple, safe and insensitive to subtle variation in operating parameters.
Amenability to safe operation in a remotely operated shielded facility with negligible operational constraints.
Flexibility to scale up or down to its level of operation in response to requirements.
Generates a minimum quantity of radioactive waste.
Offers the possibility of recovering the ytterbium for target recycling.
Taking into consideration the above-mentioned criteria, a great deal of effort has been expended upon the development of a number of Yb/Lu separation strategies. Essentially every conceivable separation approach has been profusely exploited. Methods ranging from ion exchange chromatography to electrochemical separation strategies have been employed. Brief overviews of these approaches are elaborated in the following sections.
Among the techniques used in radiochemical separation, ion-exchange chromatography has been generally proved to be a widely utilized, reliable, and straightforward way to separate radionuclides of interest for myriad applications. While the ion-exchange chromatography technique is appealing in terms of operation simplicity and amenability to a remotely operated facility, a straightforword separation of Yb and Lu is a difficult task owing to their striking similarities in chemical properties. Following a somewhat different path is required in which both Yb and Lu can be adsorbed on a cation exchanger and elution by an appropriated complexing agent. In this premise, two equilibria are required to be considered, i.e., the equilibrium between the complexing agent and the ion exchanger and the equilibrium between the Yb and Lu and the complexing agent. The difference in stability constants for the Yb and Lu with complexing agents forms the genesis of separation. The order of elution of Yb and Lu as well the resolution of the elution band depends on the stability constant values of the formed complexes. Since the smaller ions show a greater preference for complexation, Luaq3+ is the first to emerge from the column, followed by Yb. While the well-characterized α-hydroxyisobutyrate (α-HIBA) complexant as an eluting agent is useful for the separation of Lu from Yb, a Lu/Yb separation factor (α) of only 1.55 [59, 60] has been a major limitation. Owing to the low separation factor, the lutetium fraction contains significant levels of ytterbium because of the “peak tailing.” Moreover, the α-HIBA complex of 177Lu is not optimally suited for the routine synthesis of 177Lu-labeled radiopharmaceuticals. In light of the explicit need to use 177Lu for the preparation of radiopharmaceuticals, α-HIBA must be decomposed and removed prior to labeling because of its high stability constant. The transfer of 177Lu out of the thermodynamically very stable 177Lu-α-HIBA species prior to labeling constitutes a necessity as its presence not only leads to poor labeling yield, but also requires post-labeling purification. In an attempt to circumvent this drawback, one of the methods used for the removal of α-HIBA is the adsorption on a cation exchanger followed by elution with ~9 M HCl . Use of ethylene-diamine-tetra-acetate or 1,2-diamino-cyclohexanetetraacetate (α = 1.7) in lieu of α-HIBA met with limited success because of solubility problems and the requirement of additional steps to obtain 177Lu of desired purity amenable for the preparation of radiopharmaceuticals . Despite these inherent shortcomings, interestingly and surprisingly enough, the enthusiasm for using the ion-exchange chromatography technique has resulted in some investigators evaluating this pathway.
Balasubramanian reported  the separation of 177Lu from 10.35 mg of neutron-irradiated ytterbium using Dowex 50 W × 8 (200–400 mesh), a cation exchanger in Zn2+ form. 177Lu was eluted using 0.04 M α-hydroxyisobutyric acid at pH 4.6 ± 2 at 26 ± 1 °C in about 4 h. Lutetium-177 was separated in 70 % yield and with a radionuclidic purity greater than 99 %. While the reported method was successful in isolating 177Lu, more than 30 % of 177Lu contaminated with ytterbium was sacrificed. Lutetium-177 obtained from this method has low specific volume and contains the barrier-ion Zn2+. The presence of Zn2+ in the eluate is a major obstacle in the complexation chemistry of 177Lu and therefore necessitates purification as well as concentration prior to labeling.
On a similar theme, Hashimoto et al. reported the utility of reversed-phase ion-pair chromatography using a Resolve C18 column in which 177Lu was eluted with a mixture of 0.25 M α-HIBA as a complexing agent and 0.1 M 1-octanesulfonate as an ion-pairing agent . In this procedure, 5 mg of the neutron-irradiated Yb2O3 target was used, and the process was effective to provide radiochemically pure 177Lu with 84 % yield. Although this method was productive with small amounts of Yb2O3 target (0.01-1 mg), the separation efficiency deteriorated when higher amounts of the Yb2O3 target were used, resulting contamination from distortion of the ytterbium peak tailing into the lutetium peak.
Liquid-liquid extraction is one of the most promising techniques often used for radiochemical separation. Significant practical experience has accumulated over the years in using this technique in a highly radioactive environment and on an industrial scale. Although use of the liquid-liquid extraction method based on the different extractability of Lu and Yb acidic organ phosphorus extractants holds promise, the requirement of a multistage process, which is essential to achieve the necessary decontamination of Yb from Lu owing to the low Lu/Yb separation factor (α), constitutes the major impediment that probably limits its wide-scale utility. The liquid cation exchanger, di-(2-ethylhexyl)phosphoric acid (HDEHP), has effectively been utilized as an extractant for the isolation of 177Lu in proton-activated ytterbium . In this method, 0.2 g of proton-irradiated nat Yb2O3 target was dissolved in 1 M HCl, and the 177Lu formed was extracted in the organic phase containing 1 % HDEHP in cyclohexane. It is apparent that liquid-liquid extraction for 177Lu separation is still in its infancy and presently represents a potential separation technique, but much additional effort is required in order to realize its potential.
The supported liquid membrane (SLM) method for Lu/Yb separation has its roots in the liquid-liquid extraction method in which a Lu-selective organic extractant is impregnated on an inert semipermeable membrane, and separation of Lu is achieved by its selective transport through the pores of the impregnated membrane. In order to tap the potential of SLM for the radiochemical separation of 177Lu from Yb, HDEHP in hexane was impregnated on a membrane consisting of two blocks (one made of PVDF and the other of PTFE) with identical channels of dimensions. The membrane thickness was 200 μm, and its nominal pore size was 0.2 μm. The donor side of the membrane contained 0.2 mol dm−3 of ammonium acetate buffer at pH 5–5.5 in which the neutron-irradiated Yb target solution was added and the acceptor side contained 2 mol dm−3 HCl in which 177Lu was collected . Despite promising results, this separation procedure has never been extended for the separation of 177Lu from neutron-irradiated Yb. The scale of 177Lu separation possible by this route will be limited but could still be of interest and utility in meeting local needs. Continuing research in this separation methodology can be expected in the near future.
An alternative to liquid-liquid extraction is the possibility to incorporate an extractant or a solution of an extractant into an inert substrate that can be used as a support in a column chromatographic technique. The most striking feature of the extraction chromatography (EXC) technique is that it combines the selectivity of liquid-liquid extraction with the ease of operation and rapidity of a column-based separation system. It is critical, however, that an appropriate extractant needs to be chosen that offers a satisfactory Lu/Yb separation factor (α).
The EXC technique has been explored by Knapp et al., leading to the development of a one-step extraction chromatography separation process [56, 57, 67] and making use of the commercially available LN Resin, which comprises di(2-ethyl-hexyl) ortho-phosphoric acid (HDEHP), commercially available from Eichrom Technologies, Inc. The reported method was found to be effective for the quantitative separation of 177Lu from 10 mg of nonradioactive Yb carrier using HCl of different concentrations for sequential elution of 170Tm 176Yb and 177Lu. The elution sequence consisted of an initial elution with 2 M HCl (fraction 1) followed by increasing the acid concentration to 3 M HCl (fraction 2) and then 6 M HCl (fraction 2). The first peak (fraction 1) in the chromatogram contained 170Tm formed by neutron activation of a stable 169Tm impurity in the enriched 176Yb target material. The ytterbium peak (fraction 2) then appeared and finally the 177Lu was eluted with 6 M HCl. The specific activity of 177Lu obtained by this method was estimated to be 3.7 TBq (100 Ci)/mg (i.e., 91 % of the 110 Ci/mg theoretical value).
The aforementioned EXC technique was further exploited by Horwitz et al. and culminated in a conceptual flow sheet that was found to be successful for the separation of NCA 177Lu from a 300-mg irradiated ytterbium target . The process is essentially based on the use of two different EXC resins, namely a resin containing HEH[EHP] (LN2) and a resin containing tetraoctyl diglycolamide (DGA) sorbed onto Amberchrom® CG-71. The whole separation process can be broadly divided into three steps: (1) the front-end target removal step, (2) primary separation step and (3) secondary separation step. While the goals of each separation step differ, it basically consists of separation of Yb and Lu using the LN2 resin followed by the concentration and acid adjustment of the Lu-rich eluate using Amberchrom® CG-71 resin. The use of the diglycolamide EXC material to purify the Lu-rich eluate is the novelty of this technique. Using Amberchrom® CG-71 resin seemed attractive as it precludes lengthy evaporations and acidity adjustments between successive LN2 resin column runs and at the same time is effective in removing adventitious metal ion impurities from the 177Lu fraction. It is worth mentioning that during the purification of the 177Lu fraction by LN2 resin, metal ions such as Zr4+, and Hf3+ (177Hf is the daughter of 177Lu) are strongly retained and therefore free 177Lu from metallic impurities. With a view to eliminating all traces of nitrate ions, a small anion-exchange column in the final step of the secondary separation step has been added. All the 176Yb fractions of the target removal step, primary separation step and secondary separation step were pooled together and could be used for recycling in successive neutron irradiations.
A notable feature of this method is thus the recovery of the isotopically enriched 176Yb target material. The individual decontamination factors for the front-end target removal system, primary separation system and secondary separation system are 101, 102 and 103, respectively. The overall recovery of 177Lu was reported to be 73 %. The total processing time employing the three steps was reported to be 4 h. The simplified flow sheet of the front-end target removal step, primary separation step and secondary separation step are depicted in Figs. 6, ,77 and and8,8, respectively. This method is attractive owing to the commercial availability of LN2 and Amberchrom® CG-71 resin, adaptation of the user-friendly EXC process, shorter processing time, satisfactory 177Lu yield and amenability to routine remote operation as well as automation, and it offers the potential to recover the enriched 176Yb target for recycling. The prospects of adopting such a scheme appear promising for the routine production of NCA 177Lu .
Front-end enriched 176Yb target removal step . The first step involves the separation of the enriched Yb target from 177Lu using LN2 resin, and the second step constitutes the concentration and acid adjustment of the Lu-rich eluate using a chromatography column containing DGA resin
Primary NCA 177Lu separation step . The first step involves the purification of 177Lu from micro amounts of Yb using LN2 resin, and the second step using DGA resin is for the concentration and acid adjustment of the 177Lu eluate
Secondary NCA 177Lu separation step . This process is essential for removing adventitious impurities from the 177Lu
In another independent study, a multicolumn solid-phase extraction (SPE) chromatography technique using di-(2-ethylhexyl)orthophosphoric acid (HDEHP)-impregnated, OASIS-HLB sorbent-based SPE resins (OASIS-HDEHP) was used [69, 70] for the separation of 177Lu from a 50-mg Yb target irradiated in a nuclear reactor with medium neutron flux (ϕ = 5 · 1013 n · cm−2 · s−1). The reported technique exploited the selectivity of OASIS-HDEHP resin for Lu in different concentrations of HCl solution for the consecutive loading-eluting cycles performed on different columns. The method was successful for the isolation of several hundred mCi of NCA 177Lu using a 50-mg Yb target irradiated in a medium neutron flux nuclear reactor (ϕ = 5.1013 n/cm2/sec). The overall separation could be carried out in 5-6 h.
As the name suggests, the electrochemical separation strategy exploits the difference between the standard reduction potentials of two radionuclides in an electrolytic medium to selectively deposit the radionuclide of interest under the influence of the controlled applied potential. The inherent advantages of electrochemical separation processes have been elaborately discussed in recent reviews [71, 72].
While the selective deposition of the radionuclide of interest from ionic state to metallic state under the influence of the controlled applied potential is a successful paradigm, applicability of this strategy for Lu/Yb separation is precluded owing to the deeply negative reduction potentials of lanthanides (more negative than hydrogen discharge) and difficulty in controlling their electrolytic deposition onto the solid cathode. In light of the perceived need to realize the potential for Lu/Yb separation following an electrochemical strategy, a somewhat different path is required. This alternate electrochemical path basically consists of selective reduction of Yb3+ to Yb2+ and its preferential transfer onto a mercury cathode exploiting the ability of Yb2+ to form amalgams with Hg.
This strategy seems attractive for the following reasons:
An examination of the redox potentials of the Yb and Lu indicates the possibility of Yb forming the bivalent state, whereas in the case of Lu, a stable bivalent state is unknown.
While Yb2+ is known to form an amalgam, Lu3+ cannot [73–76]. Therefore, Lu is difficult to deposit on the Hg cathode from aqueous electrolytes.
Offers the possibility of electrolytic reduction of Yb3+ to Yb2+ in a mildly acidic solution owing to its high hydrogen over-voltage. Such an attribute ensures no reoxidation of Yb2+ and offers easy handling and deposition of Yb onto Hg.
The electrochemical separation method is essentially based on the formation of the Yb amalgam by electrolysis into a mercury cathode or extraction into an amalgam aimed at its removal from the Yb-Lu mixtures. With a view to removing Yb, the potential of using Hg is enticing because of its high density, the insolubility of mercury in aqueous medium and the absence of adsorptive effects. In the quest for innovative approaches to separate Yb from other lanthanides, Marsh successfully exploited the electrochemical pathway using a mercury cathode [73–76], which represents one of the very early electrochemical separation strategies at a time when the utility of the electrochemical technique in separation science had not yet been established. This elegant separation technique was later effectively harnessed by McCoy, which paved the way for the first laboratory-scale separation of Lu from Yb  and showed the extraction was quite specific for Yb. Extending this theme, Onstott [78, 79] reexamined Yb reduction using a series of alkali metal salts and employed lithium citrate in place of potassium citrate.
These three successful preparative-scale separations of Lu from Yb using mercury cathodes have been reported in the literature and are discussed below.
In order to tap the potential of the electrochemical method, Lebedev et al. reported a method  that essentially consists of dissolution of irradiated Yb2O3 in hydrochloric acid, addition of sodium acetate to form sodium amalgam and extraction of Yb by sodium amalgam from Cl–/CH3COO– electrolytes into mercury. A series of four successive cementation steps each was performed in order to achieve a satisfactory decontamination factor. With a view to achieving the desired purity, the 177Lu precipitate containing trace amounts of ytterbium was dissolved in acid and adsorbed on a cation-exchange column from which 177Lu was selectively eluted using α-HIBA. In light of the explicit need to remove α-HIBA, the eluted 177Lu solution was then adsorbed on a cation-exchange column wherein both Lu and Yb were adsorbed and 177Lu was eluted with 9 M HCl. The recovery yield of 177Lu in this process was 75 %, and decontamination factor from ytterbium was found to be >106. While the reported method is appealing in terms of recovery yield of 177Lu and product quality, the requirement of a time-consuming, complicated process involving multiple cementation cycles together with the elaborate purification steps emerged as the major impediment that would be expected to restrict its wide-scale applicability. The logistics are expected to be unfavorable to carry out such a complicated process on a very regular basis.
In order to mitigate the limitation of this method, Bilewicz et al.  developed a method based on the reduction of Yb(III) to Yb(II) with sodium amalgam followed by removal of Yb by selective precipitation as the sulfate. The principal shortcoming of this precipitation method is that the separation is not clean and requires an additional ion exchange purification step to achieve the desired purity amenable for clinical use. While the reported method obviously holds promise, the processing is quite complex because of several factors influencing its performance and requires a purification step to achieve satisfactory purity. This separation strategy is not only user-unfriendly, but also could lead to varying consistencies of the purity as well as yield.
The electro-amalgamation approach developed by Chakravarty et al.  is based on the electrolytic reduction of Yb3+ to Yb2+ in lithium citrate medium followed by formation of Yb amalgam by electrolysis and extraction of Yb from the mercury cathode. A schematic diagram of the electrochemical setup used in this procedure is depicted in Fig. 9. In the two-cycle electrolysis, the first step is is the pre-elimination of the bulk of the Yb target mass, and the second step is the further purification of 177Lu. This process provides NCA 177Lu with acceptable purity and satisfactory separation yield (>90 %) within 3-4 h. The flow chart of this electro-amalgamation process is shown in Fig. 10. This strategy thus far has been confined to laboratory-scale investigations but could still be of interest and utility if adequate technological attention is imparted. The prospects for adopting such a scheme appear promising in the foreseeable future.
Schematic diagram of the electrochemical setup used for the production of 177Lu 
Production flow chart used for the isolation of NCA 177Lu following electrochemical separation technique 
Accelerator technologies could be used to produce small quantities of 177Lu, and although a number of routes can be explored that would be useful in basic research, these are not expected to really serve as the basis to undertake large-scale cost-effective production because of the extremely low cross sections of the reaction routes envisaged.
With a view to realizing the accelerator production of 177Lu, a number of studies concerning activation cross sections of the deuteron-induced nuclear reactions as well as excitation functions of the natYb(d,xn)177,173,172mg,171mg,170,169Lu reactions have been reported, and the following are of interest. Hermanne et al. studied the cross sections of deuteron-induced reactions on Yb targets and measured the cross sections between 3 and 20 MeV for Yb(d,xn)170Lu/171Lu/172Lu/173Lu/174Lu/177Lu, and Yb(d,xnp)169Yb/175Yb . Manenti et al. measured the activation cross sections of Yb(d,xn)169Lu/170Lu/171Lu/172Lu/173Lu/174Lu/176Lu/177Lu,Yb(d,xnp)169Yb/175Yb/177Yb reactions up to 18.18 MeV .
Tárkányi et al. performed a systematic study of the activation cross sections of deuteron-induced nuclear reactions and excitation functions of the natYb(d,xn) 177,173,172mg,171mg,170,169Lu,natYb(d,x)175,169Yb and natYb(d,x)173,172,168,167,165Tm reactions up to 40 MeV. Some of these reactions were evaluated for the first time . Although promising, substantial R&D and large resources are required for the technological development and assessment owing to the challenges associated with target preparation as well as the sustained operation of accelerators on a reliable and continuous basis.
Having reviewed in detail the direct and indirect strategies for reactor production and the various processing technologies used to obtain 177Lu, the quality evaluation of 177Lu is the next important issue with regard to providing this radioisotope for clinical use. Because of the requirements imposed by pharmaceutical legislation to ensure safety and efficacy, despite these encouraging prospects and the favorable results of the various production and processing methodologies, quality evaluation of 177Lu is, of course, a prerequisite before preparation of radiopharmaceuticals in the daily nuclear medicine routine.
Radionuclidic purity is defined as the ratio, expressed as a percentage, of the radioactivity of 177Lu to the total radioactivity content of the sample. Gamma spectroscopy using a high-purity germanium (HPGe) detector in conjunction with a multichannel analyzer (MCA) is used for routine determination of the radionuclidic purity of 177Lu. To be able to quantify the 177Lu and the possible impurities, the detector system must be properly calibrated for both energy and efficiency using either a series of standardized sources, each containing a single radionuclide, or a single calibrated source containing a radionuclide having several gamma photon of different energies (e.g., 152Eu) obtained from a National Metrology Institute (NMI) or commercial laboratories that can demonstrate measurement traceability to an NMI. Because of the requirement to maintain dead time and pile-up at acceptable levels (dead time <10 %) during measurement, it is mandatory to dilute the 177Lu sample appropriately. While gamma spectrometry constitutes a successful example of the determination of the radionuclidic purity of 177Lu, direct spectral analysis of 177mLu co-produced with 177Lu and other longer lived contaminants is not possible because of the overwhelming contribution of 177Lu. This difficulty is mitigated by keeping an aliquot of a 177Lu sample, which is allowed to decay for an appropriate time (i.e., ~60 days), and then analyzing it using the gamma spectrometric technique. Gamma photon peaks pertaining to 177mLu and other long-lived radionuclides can then be easily identified based on their characteristic gamma rays. A typical gamma spectrum of 177Lu obtained from the (n,γ) 177Lu production route immediately after radiochemical processing and after 70-day decay is shown in Fig. 11.
A typical gamma spectrum of a 177Lu sample aliquot obtained from the (n,γ) 177Lu production route recorded immediately (a) after radiochemical processing and (b) after 70-day decay
Radiochemical purity is defined as the ratio, expressed as a percentage, of the radioactivity of 177Lu present as 177LuCl3 to the total radioactivity of 177Lu present in the sample. With a view to determining the radiochemical purity of 177Lu after chemical separation, both paper chromatographies (PC) as well as high-performance liquid chromatography (HPLC) techniques are used. Paper chromatography using Whatman 3MM strips is the method most commonly used to test 177Lu for radiochemical purity. The PC method is simple, fast and inexpensive. A small aliquot (~5 μL) of the test solution can be spotted at 1.5 cm from the bottom of a paper chromatography strip. The strip needs to be eluted using 0.9 % NaCl (w/v) in 0.02 M HCl as the eluting solvent. After elution, the strip can be dried and cut into segments (i.e., typically 1 cm). The radioactivity associated with each segment can be determined by using a well-type NaI(Tl) scintillation counter by keeping the base line at 150 keV and with a window of 100 keV, thereby utilizing the 208-keV gamma photon of 177Lu. A typical paper chromatography pattern of 177Lu3+ is shown in Fig. 12. For HPLC analysis, a typical system utilizing water (A) and acetonitrile (B) mixtures with 0.1 % trifluoroacetic acid is used as the mobile phase. A typical HPLC pattern of 177Lu3+ is illustrated in Fig. 13.
Paper chromatography pattern of 177Lu3+
HPLC pattern of 177Lu3+
In light of the explicit need to perform radiolabeling with 177Lu, the chemical purity is also of paramount importance, especially for receptor-targeted agents. In view of the extremely low concentration of 177Lu, the metal ion impurities even at ppb levels act as pseudocarriers, requiring higher concentrations of the targeting vectors to achieve high radiolabeling yields. Recognizing the ability of competing metal ion impurities, such as Al, Ca, Cu, Fe, Pb and Zn, likely to be present in the 177LuCl3 solution, to form thermodynamically and kinetically stable coordination complexes with the targeting vectors, it is of utmost importance to determine their concentration, which could be effectively achieved by the inductively coupled plasma atomic emission spectrometry (ICP-AES) technique.
Another noteworthy chemical impurity is the hafnium isotope, which is produced through the decay of 177Lu, 177Hf (177Lu β−−→177Hf ). Fortunately, its presence is not of much concern owing to the negligible ingrowth and due to the fact that Hf does not interfere with the labeling .
With a view to using 177Lu for targeted radionuclide therapy, the goal of attaining the highest possible specific activity is crucial. With a view to realizing this objective, the presence of cold Lu should be minimized to the extent possible as it acts as a competitor for labeling positions on targeting vehicles. On this premise, determination of the specific activity of 177Lu prior to radiolabeling was deemed worthy of consideration.
The specific activity of 177Lu can be expressed as:
Here A177Lu is the measured activity of 177Lu at any particular point in time, and m177Lu, m175Lu, mLu are the mass of 177Lu, 175Lu and cold Lu present in the sample.
The activity of 177Lu in a given aliquot is generally measured following gamma spectroscopy using an HPGe detector. The sample can be placed for the appropriate time at a suitable geometry, and the counts acquired under 208 keV after chemical processing can be used for assay of 177Lu. The total concentration of 177Lu, 175Lu and cold Lu in the sample can be determined by the ICP-AES technique. From the determination of the activity and total Lu concentration, the specific activity of 177Lu is computed.
The current worldwide suppliers of good manufacturing practices (GMP) producing 177Lu as a radiochemical are provided in Table 4. In addition to the major producers and suppliers, some countries also produce small quantities of 177Lu for domestic use.
Current suppliers of GMP-produced 177Lu as a radiochemical
Lutetium-177 produced from any of these production routes is considered as an active pharmaceutical ingredient (API) since it is used as a starting material for the preparation of radiopharmaceuticals for human use and production must be regulated. The emphasis on quality is most prominently manifested by the fact that all equipment, instruments and technologies in 177Lu production facility and the associated accessories must meet the preset criteria and the product obtained has to meet strict specifications. Written and approved protocols specifying the critical steps and acceptance criteria must be in place. Confirmation of appropriate regulatory conditions for aseptic processing and its supportive activities is mandatory. Production of 177Lu should be carried out according to GMP, which is becoming mandatory in most countries.
In order to achieve GMP compliance, it is essential to have a full documentation system providing traceability that includes:
A Site Master File
Drug Master Files for the individual batch
Validation Master File
Specifications for materials
Batch processing records
Training of staff
The US Food and Drug Administration (FDA) approved a set of regulations describing production of radionuclides used as APIs according to cGMP, outlined in the Code of Federal Regulations. In order to address these regulatory demands, radionuclide production is migrating toward the use of automated modules. The advantages of using automated production strategies include:
Assuring reproducibility in production yield as well as consistency in product quality.
Improving the robustness of the production as well as providing on-line documentation of the process, thus improving GMP compliance.
Providing a log of the steps performed during the processing of 177Lu. Electronic record keeping is not only accurate and complete, but also helps in accomplishing regulatory compliance.
Precluding the possibility of cross-contamination.
Ability to handle multiple GBq levels of radioactivity safely, enabling the manufacturer to produce and distribute the required quantities of 177Lu for therapy.
Facilitating regulatory compliance through manufacturer installation qualification/operational, qualification/performance, qualification and scheduled maintenance protocols performed for 177Lu production by trained personnel.
Improving radiation safety through the reduction (or elimination) of manual operations.
Use of automated production strategies represents an appealing vision where significant resources and effort have been expended. While automation holds promise and offers numerous advantages, it presents radiochemists with the challenge of re-configuring the chemical processing steps that require integration of several steps while maintaining full automation. Nonetheless, to be effective in addressing the particular regulatory barriers, automated processing modules must be customized to local legislative, regulatory and institutional conditions for which a comprehensively designed and correctly implemented quality assurance system is of utmost importance.
In addition to meeting pharmaceutical GMP and gross domestic product (GDP) regulations, manufacturers undertaking regular production of 177Lu must be licensed by a Nuclear Regulatory Authority (NRA). In this context, it is mandatory for the manufacturer to demonstrate that its facility used for 177Lu production is adequate to protect health and minimize danger to life and property. Additionally, the manufacturer must be qualified to use radioactive material, establish a radiation protection program as well as the controls and procedures for the management, record keeping, accounting and use of radioactive materials.
This overview of the existing 177Lu reactor production and processing technologies along with the recent developments reveals two competitive options, each having relative advantages and disadvantages. Production of 177Lu is inextricably linked to the advancements in TRT. A wide range of innovative new targets, lead compounds and new radiolabeled ligands as vectors are emerging far more rapidly than over the past decade. As radionuclide therapy is moving to the forefront of molecular-targeted radionuclide therapy of cancer and other diseases, the demand for 177Lu is evolving.
Although the accelerator-based 177Lu production option holds promise as an innovative approach, current global trends in this production route are demonstrably unsustainable both technically and economically. Completing the technological development as well as establishing the economics of this approach is expected to be years away, and its success will depend on how these challenges are tackled in the years to come. Of the two reactor production options discussed, the prospect of using CA 177Lu produced by the “direct” route is appealing as it is the least intricate way to obtain 177Lu of reasonable specific activity and will suffice for most applications. While the “direct” production route is attractive in terms of simplicity in target processing and cost effectiveness, the burden of 177mLu in the final product is the key factor in determining its usefulness. Owing to the inherent requirement of an elaborate intricate radiochemical processing technology for the isolation of NCA 177Lu, significant expertise, skilled technicians and adequate resources for undertaking regular production, the number of commercial radioisotope suppliers of NCA 177Lu remains finite, and its current production capabilities are still limited.
In recent years, targeted radionuclide therapy has been moving from an exotic treatment modality for a very few patients to a mainstream modality. The future of targeted radionuclide therapy is, of course, difficult to predict, and there will be surprising inventions that, as in the past, may have an unexpected application that will continue to fuel the field. These represent the niche areas where NCA 177Lu will have an advantage over CA 177Lu. While undertaking large-scale 177Lu production, it is essential to assess both options, weigh pros and cons, and select the one based on the technical and economic resources. It is important to note that these two reactor production routes should not be approached as competitive, but instead provide 177Lu for a variety of clinical applications to benefit needy patients.
Lutetium-177 seems destined to find important applications in the personalized therapy of patients using low-abundance gamma photons for diagnosis. This paradigm, when properly enforced, would not only provide a clear understanding of the disease following its detection and progression, but also provide vital clues for making decisions about individualized treatments. Administering suitable 177Lu-labeled radiopharmaceuticals in their required doses and providing personalized treatment planning constitute a major step forward to meet the challenges of personalized medicine. Implementation of this regimen is likely to trigger profound structural changes in the treatment strategy and potentially to create a situation where treatments can be tailored to individual patient-specific diseases. Effective harnessing of such a treatment regimen requires a constant and reliable supply of 177Lu of the required quality in the desired quantities at a reasonable cost. Because of the pace at which the personalized therapy scene is evolving, 177Lu production strategies need a vision for today and tomorrow.
The advances made in large-scale 177Lu production so far are exciting, and there are no apparent barriers to its adoption for large-scale production. With the appropriate selection of a production route, it would be possible to envision a future where the scale and potential of 177Lu production technology can be tailored to institutional needs. The progressive fusion of existing 177Lu production technologies with automation can be consciously nurtured in effective ways to respond to GMP compliance and to surmount regulatory barriers. Potentially the infusion of automation into 177Lu production technology may be hastened by the creation of a positive platform for future growth. Interest in and expansion of 177Lu production and processing technologies as well as the development and clinical introduction of 177Lu-based therapeutic radiopharmaceuticals have passed many milestones, and it is expected that broader use and regulatory approval of 177Lu-agents will move forward rapidly.
Ashutosh Dash, Maroor Raghavan Ambikalmajan Pillai and Furn F. (Russ) Knapp, Jr., declare that they have no conflict of interest.
The manuscript does not contain clinical studies. There is no identifiable patient information in this manuscript.
The information comes from:https://www.ncbi.nlm.nih.gov/pmc/articles/PMC4463871/
Purpose of the Study: With rapid development in the field of nuclear medicine therapy, radiation safety of the personnel involved in synthesis of radiopharmaceuticals has become imperative. Few studies have been done on estimating the radiation exposure of personnel involved in the radio labeling of 177Lu-compounds in western countries. However, data from the Indian subcontinent are limited. We have estimated whole body radiation exposure to the radiopharmacist involved in the labeling of: 177Lu-DOTATATE, 177Lu-PSMA-617, and 177Lu-EDTMP. Materials and Methods: Background radiation was measured by keeping a pocket dosimeter around the workbench when no radioactive work was conducted. The same pocket dosimeter was given to the radiopharmacist performing the labeling of 177Lu-compounds. All radiopharmaceuticals were synthesized by the same radiopharmacist with 3, 1 and 3 year experience, respectively, in radiolabeling the above compounds. Results: One Curie (1 Ci) of 177Lu was received fortnightly by our department. Data were collected for 12 syntheses of 177Lu-DOTATATE, 8 syntheses of 177Lu-PSMA-617, and 3 syntheses of 177Lu-EDTMP. Mean time required to complete the synthesis was 0.81, 0.65, and 0.58 h, respectively. Mean whole body radiation exposure was 0.023 ± 0.01 mSv, 0.01 ± 0.002 mSv, and 0.002 ± 0.0006 mSv, respectively. Overall mean radiation dose for all the three 177Lu-compounds was 0.014 mSv. Highest exposure was obtained during the synthesis of 177Lu-DOTATATE. Conclusion: Our data suggest that the manual radiolabeling of 177Lu compounds is safe, and the whole body radiation exposure to the involved personnel is well within prescribed limits.
Keywords: Personnel dosimetry, manual radiolabeling, radionuclide therapy
Personnel monitoring is an intergral part of any radiation safety program. Personnel monitoring aims to keep the occupational radiation exposure as low as reasonably achievable (ALARA) and is based on the principle that the benefits of any intentional or planned exposure to radiation should outweigh the resultant detriment that could arise. Safe radiation work practices and permissible radiation exposure limits have been laid by various national and international regulatory authorities.As per ICRP recommendations 103 (2007), the equivalent radiation dose to personnel should not exceed 20 mSv/year averaged over 5 years, not exceeding 50 mSv in any year. These limits are aimed at keeping the probability of stochastic effects of radiation to the lowest, while avoiding the occurrence of non-stochastic effects altogether. By defination, any person handling radiation and likely to receive an occupational radiation exposure of more than 1 mSv is liable to be monitored.In nuclear medicine, personnel involved in synthesis of radiopharmaceuticals, dose administration, and/or scan acquisition are most likely to receive radiation exposure. The risk could be even higher while handling therapeutic radiopharmaceuticals. In the present study, we have focussed on the personnel involved in synthesis of radiopharmaceuticals involving Lu-177 that is DOTATATE/DOTANOC, PSMA-617 and EDTMP. The choice of radiopharmaceuticals was based on the fact that these three radiopharmaceuticals are being routinely synthesized at our department, at the All India Institute of Medical Sciences, New Delhi, India.Over the last decade, Lu-177 has become the radionuclide of choice for various radionuclide therapy procedures owing to its ease of large-scale production in moderate flux reactors, favorable radiation characteristics enabling imaging along with therapy (β-max: 497 keV; γ1: 113 keV; 6.4% and γ2: 208 keV; 11%); and sufficiently long half-life (6.7 days) allowing easy transport to centers far off from a reactor site. These economic, characteristic, and logistic advantages of Lu-177 have become even more significant in developing countries, where affordable therapeutic options are always sought.Although there is a plethora of literature on the internal dosimetry or patient dosimetry with 177Lu-radiopharmceuticals, there is a lesser literature on the personnel dosimetry, especially those involved in synthesis. 177Lu-DOATATATE/DOTANOC, PSMA-617, and EDTMP can be synthesized in automatic or semi-automatic chemistry modules or by manual methods. Since a manual method is more cost effective than automatic or semi-automatic methods, it is the most widely practiced method in developing countries like India. However, it poses a risk of comparatively higher radiation exposure to the personnel involved. Therefore, the present study aims to monitor the radiation dose levels to personnel during manual synthesis of 177Lu-labeled compounds (DOTATATE/DOTANOC, PSMA-617, and EDTMP) and reviews work practices that may reduce the radiation exposure.
Lu-177 as LuCl3 was procured from BRIT, Mumbai, India. A digital pocket dosimeter (MyDose Mini) was obtained from ALOKA. The precursors used in the synthesis of 177Lu-labeled DOTATATE/DOTANOC and PSMA-617 were obtained from ABX GmbH, Germany, and EDTMP kit was obtained from BRIT/Polatom. All other reagents used in labeling were of analytical grade.ProcedureSynthesis of Lu-177-labeled DOTATATE, PSMA-617, and EDTMP was carried out by designated skilled personnel at the radio-pharmacy laboratory of the Department of Nuclear Medicine, AIIMS, New Delhi, India. These radiopharmaceuticals were routinely synthesized in our department, on fortnightly basis by manual methods. DOTATATE and DOTANOC were labeled alternatively depending on the availability of precursor. The MyDose mini radiation pocket dosimeter was used to measure the radiation exposure. Initially, background radiation of the laboratory, where labeling was carried out, was measured by placing the dosimeter in the laboratory when no radioactive work was being conducted. The background exposure readings were taken at different places around the labeling workbench and mean was calculated.Personnel were issued a pocket dosimeter prior to the start of labeling procedure. Initial reading of the meter was set at zero every time. Radiation exposure readings recorded in the meter were noted on the completion of labeling process. The total amount of radioactivity handled during labeling and the duration of each labeling procedure were noted.Statistical AnalysisDescriptive statistic analysis was done for the collected data; and mean, median, standard deviation (SD), and range (minimum to maximum value) were determined. All the readings were expressed as mean ± SD.
A total of 23 readings of radiation exposure were obtained during the labeling of all 177Lu-radiopharmaceuticals put together. [Table 1] shows the number of readings obtained for individual radiopharmaceuticals. Background radiation exposure reading was observed to be zero (for 1 hour) around the labeling workbench when no radioactive work was being conducted.
The details of radiation dose during labeling of 177Lu-DOTATATE/NOC, PSMA-617, and EDTMP are given in [Table 2], [Table 3], and [Table 4], respectively. [Figure 1] represents the trend of radiation exposure during labeling of the three radiopharmaceuticals.
The mean radiation dose recorded in 177Lu-DOTATATE/NOC labeling was 0.023 ± 0.01 mSv, 177Lu-PSMA-617 was 0.01±0.002, mSv and 177Lu-EDTMP was 0.002 ± 0.0006 mSv and the mean duration of labeling was 0.81, 0.65, and 0.58 h, respectively. The specific activity of Lu-177 was ~19–22 mCi/µgm in all labeling procedures.The mean estimated radiation dose rate during the three labeling procedures was 0.03 ± 0.01 mSv/h for DOTATATE/NOC, 0.01 ± 0.003 mSv/h for PSMA-617, and 0.003 ± 0.001 mSv/h for EDTMP. Overall mean radiation dose was 0.014 mSv and duration was 0.72 h.
The objective of the study was to evaluate the radiation dose levels to personnel involved in the labeling of 177Lu-labeled radiopharmaceuticals that is DOTATATE/NOC, PSMA-617, and EDTMP. The method of labeling these compounds with Lu-177 may be automated/semi-automated, or manual. At our department, we perform routine radiolabeling of these compounds with Lu-177 by a manual method, as it is more cost effective and automated modules are not available to us at present. However, in manual labeling procedures, the radiation safety concerns are higher than that in automated or semi-automated methods.Labeling of 177Lu-DOTATATE/NOC was performed as per the method described by Das et al. and that of 177Lu-PSMA-617 was performed by the method described by Ahmadzadehfar et al. Automated or semi-automated modules are not available for the labeling of 177Lu-EDTMP, as it is a single step procedure that involves simple addition and incubation of the EDTMP.Our results suggest that the labeling of 177Lu-DOTATATE/NOC yielded the highest mean radiation dose of 0.023 ± 0.01 mSv, followed by 177Lu-PSMA-617 0.01 ± 0.002 mSv, whereas the dose from the labeling of 177Lu-EDTMP was the lowest 0.002 ± 0.0006 mSv. The reason for the observed trend is the time of radiolabeling, higher the duration of radioactivity handling, higher the radiation exposure. Overall dose trend also follows the same order as can be seen in [Figure 1].One of the most important factors that may affect the radiation dose to personnel in manual methods of radiolabeling is the skill. Different radiation workers have different levels of proficiency and expertise in the handling of radioisotopes that cause the readings to vary greatly among personnel. In our study, we ensured that every time the same radiation worker was involved in the radiolabeling of a particular compound to minimize such inter-personnel variations. However, intra-personnel variations still exist. Furthermore, to maintain uniformity of measurements and minimize the errors, the pocket dosimeter assigned to particular personnel during labeling was kept the same. It was also ensured that through out the observation period (labeling process), the personnel do not carry out any other radiation work or go to any other radiation area that might yield erroneously high reading on the dosimeter.Other factors that might affect radiation dose are the duration and amount of radioactivity handled during the labeling procedure. Both, radioactivity and mean duration are highest for DOTATATE/NOC (896 mCi; 0.81 h), followed by PSMA-617 (190 mCi; 0.65 h), and EDTMP (57 mCi; 0.58 h) in our study. This explains the trend of radiation dose for the three procedures.Overall mean radiation dose for all the three 177Lu-compounds was 0.014 mSv. Our department has a high throughput of patients, and synthesis of 177Lu-compounds is performed once every fortnight, provided there is timely availability of Lu-177 and precursors. Assuming 24 such synthesis every year, the total mean dose to the personnel involved will be ~0.34 mSv. This dose level is far less than the stipulated limit of 20 mSv. Even if in future the synthesis rate increases to once per day and the same radiopharmacist is involved in synthesis, the dose will be ~5.26 mSv. The background activity in the radiolabelling laboratory returned to that existed pre-labeling, that is, zero (for 1 hour) after proper disposal of radioactive vials, syringes, absorbent sheets, gloves, and other contaminated waste. These things were properly sealed, labeled, and stored in a waste disposal room for decay. The reading of the TLD badge of the personnel involved was also within prescribed limits, that is, 0.9 mSv for chest badge for 1 year. It should be noted that this reading includes the radiation expsure to the presonnel from other sources as well apart from the radiolabeling procedures mentioned in this study as the personnel was involved in other departmental work also. This shows that even the manual radio-labeling methods of Lu-177 compounds are safe, provided safe work practices are followed.The dose can be further reduced by involving staff well trained in good radio-pharmacy practices and radiation safety. Though the procedures are safe even if a single trained staff member conducts all the synthesis, it would be preferable to involve minimum two trained personnel to share and further reduce the radiation burden. The regular use of radiation monitoring devices such as the pocket dosimeters and TLD badges should be encouraged, and radiation surveys should be routinely conducted.The study was conducted over a period of 6 months and various logistic reasons such as unexpected delay in delivery of Lu-177, or precursors sometimes restricted the regular synthesis of 177Lu-compounds at our department. Hence, not much data points could be collected that is a major limitation of the study. Furthermore, due to unavailability of automated/semi-automated chemistry modules at our department a direct comparison was not possible. However, despite a less number of observations, the study is significant as there are only few similar studies on radiation dose levels to personnel involved in Lu-177 radio-labeling.
Our data suggest that the manual radio-labeling of 177Lu-compounds is safe and the whole body radiation exposure to the involved personnel is well within the prescribed limits of ICRP, i.e., 20 mSv/year (averaged over 5 years). However, the exposure can further be reduced using semi-automated and automated modules, wherever possible.
The information comes from:http://www.ijnm.in/article.asp?issn=0972-3919;year=2017;volume=32;issue=2;spage=89;epage=92;aulast=Arora;type=3
(99m)Tc-MDP (technetium-99(m)-labeled methylene diphosphonate) has been widely used as a radiopharmaceutical for bone scintigraphy in cases of metastatic bone disease. (177)Lu is presently considered as an excellent radionuclide for developing bone pain palliation agents. No study on preparing a complex of (177)Lu with MDP has been reported yet. Based on these facts, it was hypothesized that a bone-seeking (177)Lu-MDP (lutetium-177-labeled MDP) radiopharmaceutical could be developed as an agent for palliative radiotherapy of bone pain due to skeletal metastases. Biodistribution studies after intravenous injection of (177)Lu-MDP complex in rats may yield important information to assess its potential for clinical use as a bone pain palliation agent for the treatment of bone metastases.
(177)Lu was produced by irradiating natural Lu(2)O(3) (10 mg) target at a thermal flux ∼ 8.0 × 10(13) n/cm(2) per second for 12 h in the swimming pool-type reactor.(177)Lu was labeled with MDP by adding nearly 37 MBq (1.0 mCi) of (177)LuCl(3) to a vial containing 10 mg MDP. The radiochemical purity and labeling efficiencies were determined by thin layer chromatography. Labeling of (177)Lu with MDP was optimized, and one sample was subjected to high-performance liquid chromatography (HPLC) analysis. Twelve Sprague-Dawley rats were injected with 18.5 MBq (0.5 mCi). (177)Lu-MDP in a volume of 0.1 ml was injected intravenously and then sacrificed at 2 min, 1 h, 2 h and 22 h (three rats at each time point) after injection. Samples of various organs were separated, weighed and measured for radioactivity and expressed as percent uptake of injected dose per gram. Bioevaluation studies with rats under gamma-camera were also performed to verify the results.
The quality control using thin layer chromatography has shown >99% radiochemical purity of (177)Lu-MDP complex. Chromatography with Whatman 3MM paper showed maximum labeling at pH = 6, incubation time = 30 min, and ligand/metal ratio = 60:1. HPLC analysis showed 1.35 ± 0.05 min retention time of (177)Lu-MDP complex. No decrease in labeling was observed at higher temperatures, and the stability of the complex was found adequate. Biodistribution studies of (177)Lu-MDP revealed high skeletal uptake, i.e., 31.29 ± 1.27% of the injected dose at 22 h post injection. Gamma-camera images of (177)Lu-MDP in Sprague-Dawley rats also showed high skeletal uptake and verified the results.
The study demonstrated that MDP could be labeled with (177)Lu with high radiochemical yields (>99%). The in vitro stability of the complex was found adequate. Biodistribution studies in Sprague-Dawley rats indicated selective bone accumulation, relatively low uptake in soft tissue (except liver) and higher skeletal uptake, suggesting that it may be useful as a bone pain palliation agent for the treatment of bone metastases.
The information comes from:https://www.ncbi.nlm.nih.gov/pubmed/21492790
Rationale The image-guided iodine-125 seed implantation has been widely used for a variety of tumors, including prostatic cancer, pulmonary cancer, hepatocellular carcinoma and pancreatic cancer. However, the clinical value of iodine-125 seed implantation for the treatment of pancreatic metastasis from hepatocellular carcinoma has not been reported. We presented the first case with ultrasound-guided iodine-125 seed implantation for this disease. Patient concerns We presented the case of a 48-year-old man patient with primary hepatocellular carcinoma and pancreatic metastasis who was managed with ultrasound-guided iodine-125 seeds implantation. Diagnoses She was diagnosed with synchronous pancreatic metastases from hepatocellular carcinoma. Interventions Puncture biopsy and ultrasound-guided iodine-125 seeds implantation. Outcomes The hepatic and pancreatic tumors were obviously reduced after 15 months. Moreover, the liver function test was mildly abnormal in glutamic-oxalacetic transaminase and glutamic-pyruvic transaminase. Lessons The image-guided iodine-125 seeds implantation was an important therapeutic approache to unresectable hepatocellular carcinoma with pancreatic metastasis. However, more related cases should be reported for further evaluating the value of the way.
The information comes from:https://www.researchgate.net/......
Brachytherapy for prostate cancer by means of permanently implanted 125I sources is a well established procedure. An increasing number of patients all over the world are treated with this modality. When the technique was introduced at our institution, radiation protection issues relative to this technique were investigated in order to comply with international recommendations and national regulations.
Particular attention was paid to the need for patient shielding after discharge from hospital. The effective and equivalent doses to personnel related to implantation, the effective dose to patient relatives as computed by a developed algorithm, the air kerma strength values for the radioactive sources certified by the manufacturer compared with those measured by a well chamber, and the effectiveness of lead gloves in shielding the hands were evaluated. The effective dose to the bodies of personnel protected by a lead apron proved to be negligible.
The mean equivalent doses to the physician's hands was 420 microSv for one implant; the technician's hands received 65 microSv. The mean air kerma rate measured at the anterior skin surface of the patient who had received an implant was 55 microGy/h (range, 10-115) and was negligible with lead protection.
The measured and certified air kerma strength for125I seeds in RAPID Strand corresponded within a margin of +/- 5%. The measured attenuation by lead gloves in operative conditions was about 80%. We also defined the recommendations to be given to the patient at discharge. The exposure risks related to brachytherapy with 125I to operators and public are limited. However, alternation of operators should be considered to minimize exposure. Patient-related measurements should verify the dose rate around the patient to evaluate the need for shielding and to define appropriate radiation protection recommendations.
The information comes from:https://www.researchgate.net/publication/7491399_Prostate_brachytherapy_with_iodine-125_seeds_Radiation_protection_issues
James Nairne,... Andreas Meijer, in Progress in Medicinal Chemistry, 2015 3.1.1 Technetium-99m Tracers Technetium-99m is produced relatively inexpensively using a generator. Molybdenum-99 suspended on an alumina column decays (t½ = 66 h) to form technetium-99m. The singly charged 99mTcO4− is eluted in preference to the doubly charged 99MoO42− using saline. Commercially available technetium-99m radiotracers are generally prepared by the simple addition of technetium-99m eluted from the generator to a kit vial containing a freeze-dried formulation of the active ingredient. The technetium-99m half-life of 6 h allows time for preparation of the radiotracer, distribution and patient imaging. The energy of the γ-ray emission (140 keV) is ideal for imaging using gamma cameras. Technetium has a rich coordination chemistry with several potential oxidation states . Most nuclear imaging agents contain technetium-99m in the + 5, + 3 or + 1 oxidation states, although it is also present in the + 7 oxidation state in the thyroid imaging agent 99mTcO4−, as formed in the generator. Technetium has good affinity for nitrogen, oxygen, phosphorus and sulphur in the most common oxidation states. The preparation of technetium-99m imaging agents is relatively straightforward (Scheme 1); a kit comprising a reducing agent, usually stannous chloride, a weak chelating agent and the cheland is treated with the generator eluate and the mixture incubated for a short time, often at room temperature, giving a preparation that is ready for injection without purification.
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The information comes from:https://www.sciencedirect.com/topics/neuroscience/technetium-99m
60 years ago in the Brookhaven National Laboratory (BNL), the DOE's cutting-edge nuclear research facility located on Long Island, the invention of a new isotope production method gave birth to the field of nuclear medicine.
Developed by radiochemists Walt Tucker, Powell Richards, and Margaret Greene, the seemingly ordinary column-shaped device, nicknamed the " moly cow", is capable of producing a highly useful medical isotope on demand.
First discovered in the 1930s, technetium is the lightest element whose isotopes are all radioactive. Technetium-99m (Tc-99m), first generated through particle bombardment of molybdenum, produces moderately-energic gamma photons as it decays, which was recognized for its potential in non-invasive medical imaging.
But the half-life of the isotope is just over 6 hours, marking an insurmountable problem for transportation and supply chain management. But the invention at the BNL revolutionized how Tc-99m is produced--by simply eluting the generator with an aqueous solution, one can extract Tc-99m from the generator, where its parent isotope Molybdenum-99 (Mo-99, which has a moderate half-life) decays and replenish Tc-99m till the end of its own shelf life.
This breakthrough allowed pharmacists to generate the medical isotope on demand inside healthcare facilities, making it much easier for doctors and patients around the world to access nuclear medicine. It was estimated that over 40 million nuclear imaging procedures are done every year using the isotope, allowing doctors to conduct life-saving diagnoses.
Due to the versatility of its chemical form, scientists have designed a variety of Tc-99m-based radiopharmaceuticals that can target specific parts of the human body, therefore enabling the diagnosis of different diseases.
For instance, Tc-tetrofosmin and Tc-sestamibi are popular imaging tracers for myocardial perfusion; Tc-macroaggregated albumin is used in lung perfusion imaging and venography for detecting deep vein thrombi; Technetium phosphonate such as Tc-MDP and Tc-HDP are formulated for bone imaging. On top of chemicals, Tc-99m can also be used to label red blood cells for blood pool study.
Mid-20th century, the extensive use of Tc-99m rapidly expanded the production of its parent isotope Mo-99 to a handful of research reactors across the world, which sowed the seeds of the later shortage problem. As these reactors are getting close to the end of its intended lifespan and the attempt to bring replacement reactors online failed, the production of mo-99 was repeatedly interrupted due to extended periods of reactor maintenance in the late 2000s.
Related reading: Reactor shutdown threatens world’s medical-isotope supply
Global shortages of Tc-99m prompted researchers to look into other production methods without the use of a fission nuclear reactor. Methods such as neutron capture, proton and X-ray bombardment, and even cyclotron production have been intensively researched. But the output capacity from these alternative processes is quite limited. Meanwhile, as funding start to pour into building low-enriched uranium-fueled reactors (also known as LEUs, a result of the non-proliferating measure adopted by the international nuclear community), many foresee a future where the need of Tc-99m could be entirely covered by these new reactors.
Tc-99m was not the only star in nuclear medicine born out the Brookhaven National Laboratory. In 1976, fluorine-18 (F-18), a positron-emitting isotope was successfully incorporated into a glucose derivative, producing a metabolism tracking radiotracer known as F-18-fluorodeoxyglucose (FDG) that can be used in positron emission tomography (PET) scanning. These days, FDG is among the world’s most widely used radiotracers for cardiology, neurology, and oncology diagnosis, with more than 1.5 million 18FDG PET scans performed annually.
In the last decade, even though imaging modalities such as PET/CT, PET/MRI, and SPECT/CT are still the mainstay of nuclear medicine, the idea of a radiolabeled vehicle that's capable of diagnosis with therapy started to take off. The advent of theranostics transforms the field from "nuclear radiology" to true "nuclear medicine”.
Lutetium-177(Lu-177)-dotatate is a somatostatin analog peptide has been developed for the treatment of gastroenteropancreatic neuroendocrine tumors. Lu-177 not only delivers for precise and potent beta radiation therapy for carcinoid tumors but also emits gamma radiation that enables SPECT imaging for treatment monitoring.
Alpha particle emitting isotope actinium-225 has attracted a lot of attention due to its cytotoxic radiation. Several groups have been working on developing targeting cancer drug using this isotope. A novel optical imaging technique was also underway--the radionuclide's decay triggers Cerenkov luminescence. Capturing the optical signals from the treatment site will allow oncologists to make accurate prognoses.
Under the synergetic efforts of radiochemists, pharmaceutical scientists and radiologists, the field of nuclear medicine continues to advance and evolve. One can expect in the not-too-distant future there will be more novel radiopharmaceuticals that can meet the unmet medical needs, provide life-saving diagnosis and treatment.
The information comes from:https://www.labroots.com/trending/chemistry-and-physics/13398/nuclear-medicine-origin-crisis-renaissance
The Security of Supply Workgroup brings together the members of AIPES who operate Research Reactors or use their output for medical isotope production purposes. This involves either the production of fission isotopes (e.g. Molybdenum, Iodine, Xenon) by splitting Uranium targets, or the neutron activation of isotopes (e.g. Strontium, Iridium, Samarium, Lutetium) by bombarding stable target materials.
The main goal of this Workgroup is to achieve the best possible coordination of the periods during which the different Reactors are operating, in order to provide adequate global coverage during planned reactor shutdown periods for refuelling and maintenance. This coordination is necessary to ensure production continuity of the mother isotopes and ensure the supply of final Radiopharmaceutical products and Therapy sources to hospitals all year round. The Workgroup also provides a forum for the Research Reactor operators and the Radiopharmaceutical producers to keep each other informed about important developments in the medical isotope market and to discuss issues of common concern. The production of Molybdenum-99 (Mo-99) in particularly, is very sensitive to any unplanned unavailability of Reactors. This is because of the extremely high global demand for Mo-99 for the production of Technetium-99m Generators (Tc-99m), Tc-99m being the daughter isotope of Mo-99.
In Nuclear Medicine, around 80% of all diagnostic procedures (around 30 million per year) depend upon Tc-99m. Both Mo-99 and Tc-99m have short half lives; this excludes the possibility of building buffer stocks of these important materials.
Initially, the Workgroup concentrated on European Reactors. But following various recent events which identified that the supply situation was becoming more difficult and in view of the global scope of Mo-99 supply, it became clear that this group had to extend its sphere of activity to include all global reactor operators and distribution organizations: NRU in Canada, SAFARI in South Africa and OPAL in Australia.
In order to maximize the Workgroup’s efficiency, in 2009 the Research Reactor operators became associate members of AIPES. Therefore they are now full participants in the Workgroup meetings and this, combined with the increase in the number of participating Reactors, has contributed to a smoother and more open dialogue.
The Workgroup, assisted by the AIPES Executive Committee, has numerous contacts with European Institutions as well as with individual National Authorities. The Workgroup also cooperates with the NEA group of the OECD and supports the IAEA in global initiatives to solve the present and future Mo-99 supply problems. The Workgroup has also built close working ties with medical associations such as the European Association of Nuclear Medicine (EANM), as well as with parallel industry groups such as CORAR in the USA.
Under normal circumstances, the Workgroup meets three times a year. However, in crisis situations, meetings are more frequent and additional meetings and telephone conferences can be organised at the initiative of individual members or at the request of the Executive Committee or international governmental institutions.The Security of Supply Workgroup is committed to global coordination of Reactors and Radiopharmaceutical producers of critical medical isotopes to ensure continuity of global supply with a minimum risk of disruption. AIPES is committed to clear communication of supply related issues to the medical community and acts as the central global coordinator and communication channel at times of supply uncertainty.
The information comes from:http://www.aipes-eeig.org/spip.php?rubrique14
Although the use of Sr-89 chloride in the treatment of patients with prostate and breast cancer has been widely reported, little information is available about its use for other malignancies. Here, we retrospectively analyzed the clinical profile of Sr-89 chloride in various patients with painful bone metastases.
Entry criteria were a pathologically proven malignancy, clinically diagnosed multiple bone metastases, and adequate organ function. Sr-89 chloride (Metastron) was given by single intravenous infusion at 2 MBq/kg over 2 min. Self-reported outcome measures were used as a response index, including pain diary data on a 0–10 numeric rating scale (NRS).
Fifty-four consecutive patients with painful bone metastases were treated with Sr-89 chloride at the National Cancer Center Hospital East between March 2009 and July 2011, consisting of 26 with breast/prostate cancer and 28 with other malignancies (lung 8, head and neck 6, colorectal 6, others 8). Thirteen (24 %) patients experienced a transient increase in pain, which was categorized as a flare-up response. Grade 3–4 anemia was observed in 6 patients, 3 of whom required blood transfusion. Regarding efficacy, response rates and complete response rates were 71.2 % and 34.6 %, respectively, and time to response from the initiation of treatment was 36 days (range, 13–217). No significant difference in response rates was seen between patients with breast/prostate cancer and other cancers (breast/prostate 69.2 %, other 73.1 %; p = 0.76).
As in patients with breast and prostate cancer, Sr-89 chloride is a promising agent for the treatment of painful bone metastases in patients with various other malignancies.
The information comes from:https://link.springer.com/article/10.1007%2Fs10147-013-0597-7
This article presents a general discussion on what has been achieved so far and on the possible future developments of targeted alpha (α)-particle therapy (TAT). Clinical applications and potential benefits of TAT are addressed as well as the drawbacks, such as the limited availability of relevant radionuclides. Alpha-particles have a particular advantage in targeted therapy because of their high potency and specificity. These features are due to their densely ionizing track structure and short path length. The most important consequence, and the major difference compared with the more widely used β−-particle emitters, is that single targeted cancer cells can be killed by self-irradiation with α-particles. Several clinical trials on TAT have been reported, completed, or are on-going: four using 213Bi, two with 211At, two with 225Ac, and one with 212Pb/212Bi. Important and conceptual proof-of-principle of the therapeutic advantages of α-particle therapy has come from clinical studies with 223Ra-dichloride therapy, showing clear benefits in castration-resistant prostate cancer.
In radioimmunotherapy (RIT), monoclonal antibodies (mAb) are conjugated to radionuclides, which provide a specific internal radiotherapy. The clinical success so far has been achieved with the beta (β−)-emitting (electrons) nuclides 90Y and 131I, conjugated to anti-CD20 mAb in follicular B-cell non-Hodgkin lymphoma. The lack of success in the adjuvant setting in solid cancer (i.e., with microscopic tumor burden) may be due to the fact that emitted electrons do not deposit their main energy to the micro-metastatic tumor cells where the antibody has bound; rather, the energy (and its effects) will be released along a several millimeter long electron track, i.e., in the surrounding healthy tissue, see Figure 1.
Figure 1. The favorable geometric situation for α-particles in small-scale metastases (e.g., in the adjuvant setting) is depicted in a scanning electron microscopy micrograph of micro-metastatic clusters from ovarian cancer on the peritoneal lining (mouse). The range of the α-particles in red (here ~50–70 μm), can hardly reach the surrounding normal healthy cells other than possibly the mesothelium and its sub-layer. They cannot reach the epithelial cells of the intestinal lining. The situation for β− particles on the other hand, shows that a great deal of its energy will be deposited far away from the binding site and possibly into healthy tissue as demonstrated by the white dashed line (here ~700 μm). Consequently, it may add to side effects. Bar equals 100 μm.
This review concerns targeted alpha (α)-particle therapy (TAT), where α-emitting nuclides are conjugated to a carrier, normally an antibody. Alpha-particle decay is the release of a heavy and energetic particle, which deposits its energy in a 70–100 μm long track, i.e., within microscopic tumor cell clusters. Importantly, this high linear energy transfer (high-LET) radiation is not dependent on active cell proliferation or oxygenation, and the resulting DNA damage caused by α-particles is much more difficult to repair than that of β−. Thus, highly cytotoxic radiation directed to the relevant tumor cell deposits holds the promise of adding substantially to hitherto failing curative adjuvant chemotherapy both when administered intraperitoneally (i.p.) for ovarian cancer, and as a systemic curative adjuvant treatment for breast, colon, prostate, and other malignancies, constituting a “systemic conformal radiotherapy at the cellular level.”
Monoclonal antibodies are so far the most commonly used vector (1, 2). Other targeting agents include substrate analogs, normally in the form of peptides (3, 4), or ligands like folic acid (5). The mAb can be the whole immunoglobulin molecule or fragments like F(ab′)2 or single chain, diabodies, etc. Clearance and tumor uptake vary with size and pharmacokinetic properties, and mAb can now even be tailor-made (6).
A brief introduction to the relatively small number of early stage clinical studies using TAT in a variety of situations will follow, i.e., in recurrent brain tumor (7–9), recurrent ovarian cancer (10), human epidermal growth factor receptor-2 (HER-2) positive i.p. cancers (11), myelogenous leukemia (12–16), non-Hodgkin lymphoma (17), and metastatic melanoma (18, 19). There is also one randomized placebo-controlled trial using 223Ra-dichloride (having a high affinity for bone tissue) for symptomatic skeletal metastases in prostate cancer, the use of which is now approved by the US Food and Drug Administration (FDA) (20).
Today, the multimodal therapeutic approach often includes local gross-tumor eradication by surgery or external radiotherapy, together with or followed by regional adjuvant radiotherapy, and eventually systemic adjuvant chemotherapy. The order of these interventions may differ. As outlined, TAT is mainly aimed at microscopic residual disease and is therefore perhaps best used after adjuvant chemotherapy, but the timing and situation can vary. A number of thematic situations where TAT has, or may, be used are shortly discussed, relating both to the route of administration and/or a specific intention.
Intra-cavity administration is a natural starting point for the introduction of TAT in humans. By this approach, the risk of general side effects of critical organs, e.g., bone marrow, is minimized. Similarly, it reduces the risk of unknown toxicity due to unforeseen microscopic accumulation of the radioimmuno-complex elsewhere in the body. This relates to the use of α-particle emitters with relatively short half-life, such as 213Bi (~45 min) and 211At (~7.2 h), because most of the radioactive decay will occur within the specific cavity before the substance is distributed throughout the body via the systemic and lymphatic systems. Indeed, this has been proved in recurrent malignant gliomas and for i.p. treatment of ovarian cancer (9–11). In tumor resection cavities, the anti-tenascin mAb 211At-81C6 was administered to 18 patients with recurrent brain tumors with no grade 3 or higher toxicity, and it was concluded to be a safe treatment with some positive effects (9). With equally low toxicity, the small 11-amino acid peptide substance P (targeting the neurokinin type-1 receptor) conjugated to 213Bi has been either injected in residual tumor or in the resection cavity of glioblastoma multiforme (7, 8).
The i.p. route of administration was used in nine patients with recurrent ovarian carcinoma using 211At-MX35, an antibody against sodium-dependent phosphate transport protein 2b (NaPi2b) (10). The toxicity was mild, grade I–II, and specifically, there was no bone marrow toxicity. This was likely related to the fact that only 6% of injected initial activity concentration of the infused solution could be measured in serum, which peaked at 45 h. Additionally, 212Pb conjugated to trastuzumab, an anti-HER-2/neu receptor, for patients with HER-2 positive i.p. cancer has corroborated a low systemic distribution (11).
Adjuvant treatment for large tumor groups, e.g., breast, colorectal, and lung cancer, today includes systemically delivered chemotherapy. Although there is a clear effect on survival, in the case of colon cancer, at most, about 30% of patients harboring micrometastases are cured (21). Similarly low, or lower, figures for the total efficacy of adjuvant chemotherapy apply for breast and other adjuvant therapies. It is thought that TAT could be suitable for a boost, or consolidating, therapy after primary surgery and adjuvant chemotherapy. Besides the more common epithelial cancer where adjuvant chemotherapy is used, it has been suggested that malignant melanoma might benefit from adjuvant TAT. 213Bi-9.2.27, an antibody against human neural/glial antigen 2 (NG2), has been administered both intra-lesionally and i.v. in patients with metastatic melanoma with promising results (18, 19). The adjuvant situation is also the goal in ovarian cancer, with the benefit of using local i.p. administration (10). In future clinical trials, however, patients who would remain disease-free even without such an adjuvant therapy might be included. It will therefore be important to include stochastic and long-term risk assessments, such as secondary cancers and/or specific organ dysfunctions, in the therapy justification. In these cases, the equivalent absorbed doses in all relevant organs should be calculated, including a conservative estimate of the relative biological effectiveness (RBE) for the emitted α-particles (22).
If tumor dissemination is confined to the peritoneum today, extensive cytoreductive surgery with i.p. chemotherapy is suggested for selected patients, and i.p. TAT may be used as an additional boost therapy. An analogous local adjuvant treatment situation would be after surgery for peritoneal or pleural mesothelioma. Other multiple special-case scenarios include, e.g., optimized treatment of neuroendocrine tumors expressing somatostatin receptors, using the synthetic ligand octreotate (23), which today are treated with β−-particles such as 177Lu, if kidney toxicity could be shown to be less. In the diffuse-type gastric cancer subset, TAT using, e.g., a mutated E-cadherin mAb may represent an option for treatment (24).
Palliative treatment can be envisaged for relief of specific symptoms from localized disease using the intra-cavity route of administration like meningeal, pleural, or peritoneal carcinomatosis; the latter is currently being explored (11). Prolongation of life was found with i.v. injected 223Ra-dicloride (Xofigo®, formerly named Alpharadin) in a placebo-controlled phase III trial for castration-resistant prostate cancer metastases (20). Although 223Ra-dicloride is not conjugated to a targeting molecule, it can be considered as targeted on the basis of its affinity for bone tissue, due to similarities to calcium. The other study objectives, to give symptom relief of bone metastasis and reduce skeletal events, were also fulfilled. Hematological toxicity was surprisingly low and a good tolerability is truly important in palliative treatment. This drug is now also investigated for retreatment (25) and use in combination treatment with docetaxel (26) and also in osteosarcoma (27). A true targeted therapy (i.e., a radionuclide bound to a tumor-specific agent) in early stage prostate cancer, with only minimal metastatic disease, could be used before the appearance of bone metastasis-related symptoms. At the time when only the prostate specific antigen (PSA) level has started to increase, after optimal local and endocrine treatment, as a possible adjunct PSA salvage treatment.
Systemically dispersed myelo-lymphoproliferative malignancies are more rapidly accessible for radioconjugate binding compared with solid tumors, when considered as floating cell suspensions. However, they do form extensive aggregates in the bone marrow and in peripheral lymphoid tissues. RIT with the longer range, low energy β−-particle-emitting conjugates (Zevalin®/Bexxar®) is useful for the more bulky lymphomas and are approved for follicular B-cell non-Hodgkin lymphoma, but comes with long-lasting bone marrow toxicity (28). The safety and feasibility of TAT with 213Bi-lintuzumab (HuM195), a humanized anti-CD33 mAb that targets myeloid leukemia cells, has been established (12, 14). Importantly, anti-leukemic effects were also demonstrated, providing the first proof-of-concept in human (12). It is suggested that when introducing TAT directly after chemotherapy, the cytoreductive effect of the chemotherapy can enhance the possibility of a saturation of CD33 sites by the targeted drug, which will increase the number of radionuclides delivered to leukemia cells without the need for activity escalation (13). To even further enhance the effects, the same mAb is now being conjugated to the in vivo α-particle generator 225Ac, which decays in a serie emitting four α-particles (15), see Figure 2. Additionally, an on-going investigation is using the combination of 225Ac-lintuzumab and the cytotoxic drug cytarabine in older patients with acute myeloid leukemia (AML) (16). The surface targets used today are mostly present to a certain degree on normal hematological cells. Therefore, bone marrow toxicity is of concern and more malignant cell-specific targets are warranted.
Figure 2. Decay chains. Alpha-particle emitters are in red boxes and stable isotopes are in green boxes. The box in light green to the far right (251Cf) indicates that although the isotope is considered stable in medical applications (T1/2 = 898 years), it can still decay via 227Ac to 207Pb (stable). The T1/2 is shown inside each box, and between boxes the type of decay [α, β(−/+), or EC (electron capture)], with the probability of each decay route occurring (expressed as %). In the figure are also shown three alpha-particle emitters that are not mentioned in the text: 230U, 226Th, and 255Fm. Studies on the feasibility of producing 230U and its daughter 226Th via proton irradiation of 231Pa according to the 231Pa (p, 2n) 230U reaction have been performed (29). So far, there are no published data on the use of these three nuclides for TAT, although 255Fm has been occasionally mentioned as a potential candidate for targeted radionuclide therapy.
Regarding manifest macroscopic disease, as has been argued, this situation might not be theoretically optimal for TAT. However, there are some clinical indications that TAT may actually be of use also for treating macroscopic tumors. Firstly, there is an interesting phase I trial for manifest stage IV malignant melanoma with promising results, including an objective partial response rate of 10 and 40% of patients having stable disease at 8 weeks (19). A total of 38 patients were treated with the 9.2.27 mAb (against human melanoma chondroitin sulfate proteoglycan) conjugated to 213Bi. Secondly, preliminary reports of a phase I dose escalation trial with 213Bi-labeled anti-CD20 against relapsed or refractory non-Hodgkin lymphoma preliminary showed no acute or extramedullary toxicity in two responders out of nine treated patients (17). These results are even more promising considering the short half-life of 213Bi (~45 min), since a more long-lived nuclide would likely have been able to penetrate the tumor masses better, with possibly even better therapeutic effects. Thus, it is argued that if penetration is optimized and high enough activity is delivered to yield homogenous curative doses, also tumors in the size range of 5–10 mm can be eradicated, as has been shown experimentally (30). This potential could even be further enhanced with the use of pre-targeting strategies (see separate section).
The ovarian cancer example aims to use RIT as a locally injected adjuvant therapy. Unfortunately, epithelial ovarian cancer (EOC) mortality has not decreased during the last decades, despite a decline in incidence and treatment intensification. Diagnosis is commonly made at an advanced stage with widespread peritoneal dissemination; 70–75% of the patients are diagnosed at more advanced stages i.e., >stage I. Standard therapy for stage II and higher constitutes surgery with cytoreductive intent (i.e., removal of as much as possible of the macroscopic tumors from the peritoneal surface including bilateral salpingo-oophorectomy), supplemented by i.v. chemotherapy, and sometimes i.p. chemotherapy (31). To enhance survival, trials have assessed the use of whole abdominal or moving-strip external-beam radiotherapy (EBRT) (32), or non-specific i.p. radiotherapy with colloid preparations of 198Au or 32P as adjuvant therapies (33, 34). However, the results of these studies have not justified their routine use and long-term toxicity in normal tissues is a major concern. However, even when cytoreductive surgery and chemotherapy result in complete remission at second-look laparotomy and normalization of the serum marker cancer antigen 125 (CA-125), about 70% of patients with stage III ovarian cancer will relapse. Recurrence is often characterized by gradual development of ascites and chemotherapy-resistant tumor cells, growing as peritoneal microscopic cell deposits, eventually leading to intestinal adhesions and bowel obstruction.
Chemotherapy injected i.p. in the abdominal cavity can result in both a reduction in recurrences and a decrease in mortality, although at the cost of increased normal tissue toxicity (35, 36). The advantage of i.p. administration compared with i.v. injection for localizing radiolabeled mAb to microscopic peritoneal tumor disease was shown in earlier studies, both in animal models and in patients (37, 38). Therefore, local treatment with the β−-particle-emitting radioconjugate 90Y-HFMG (human milk fat globule-1, a mAb toward MUC-1) was investigated in a large randomized controlled phase III trial, but overall survival did not improve, although a slight decrease in local intraperitoneal recurrence was observed (39, 40). This negative result might be in part explained by the delivery of a too low absorbed dose from the emitted β−-particles to single tumor cells or micrometastases. Consequently, i.p. TAT using specific mAb labeled with α-particle-emitting radionuclides, with the higher LET and shorter path length than β−-particles, could be more effective. A phase I study has used the mAb MX35 F(ab′)2 fragments labeled with 211At, that was administered as i.p. infusion to patients with relapsed ovarian cancer but after having achieved a complete macroscopic response on second-line chemotherapy (10). The tolerability was very good and it was concluded that this treatment could achieve therapeutic absorbed doses in microscopic tumor nodules without causing any radiation-related toxicity (10).
Some important physical characteristics of relevant α-particle emitters are presented below, with reference to studies on their therapeutic applications. See Figure 2 for a schematic of the different decay pathways. Importantly, as it is not possible to directly measure the α decay in vivo, even a small amount of accompanying γ-radiation will enable scintigraphic evaluation for pharmacokinetic and dosimetric studies to be performed. All α-particle emitters with a serial decay that includes α-particle daughters can present problems, as the daughters will detach from the targeting vector due to the elevated recoil energy (up to 200 keV). Such free nuclides can then diffuse away, leading to untargeted irradiation of normal tissues. Using microdosimetry, the energy deposited in the target could be reduced by 50%, as has been calculated for the 211At α-particle-emitting daughter 210Po, with a T1/2 of 0.5 s (41).
Actinium-225 (225Ac) has a T1/2 of 10 days, causing the emission of four α-particles in a serial decay. The decay is accompanied by γ-radiation. This nuclide can have great therapeutic potential when radiochemistry can produce stable binding to 225Ac and its daughters. This nuclide is available as a consequence of producing 233U via the nuclear reaction 232Th (n, γ) 233Th (β−) 233Pa (β−) 233U for nuclear energy and nuclear weapons purposes decades ago (Figure 2). The possibility of producing 225Ac by use of a cyclotron via the 226Ra (p, 2n) 225Ac is now also investigated (42). 225Ac is currently tested in two clinical studies where it is conjugated to the anti-CD33 mAb HuM195 (15, 16).
Radium-223 (223Ra) has a T1/2 of 11.4 days and emits four α- and two β−-particles in the decay chain as well as γ-rays, until the stable isotope 207Pb is obtained. This nuclide can be produced by neutron activation of 226Ra by the nuclear reaction 226Ra (n, γ) 227Ra (β−) 227Ac (Figure 2). 223Ra is an alkaline earth metal ion and similarly to calcium ions, it accumulates in the bone. To this aim, 223Ra-dichloride was developed and is now FDA-approved for bone metastases in castration-resistant prostate cancer (20).
Bismuth-213 (213Bi) decays with a T1/2 of 45.6 min to 209Bi (stable), during which it emits one α-particle and an accompanied 440 keV γ-radiation. This nuclide can be obtained by elution of the 225Ac/213Bi generator, thereby making availability and dispersion to clinical centers possible. The generator is produced by the Oak Ridge National Laboratory in the USA and by the Institute for Transuranium Elements in Karlsruhe, Europe. Although the drawback of its short half-time puts high demand on the logistics for radiochemistry and treatment, 213Bi has still been the most used TAT nuclide in clinical trials so far (12–14, 17–19).
Bismuth-212 (212Bi) has a T1/2 of 60.6 min and emits one α- and one β−-particle. High energy (2.6 MeV) γ-rays are emitted in the decay; therefore, patients must be treated using special radiation protection routines. This nuclide is available as a consequence of producing 233U via the nuclear reaction 232Th (n,γ) 233Th (β−) 233Pa (β−) 233U (n,2n) 232U for nuclear energy and nuclear weapon purposes decades ago (Figure 2). The last step in which 232U was produced via the (n, 2n) reaction was an unwanted side reaction during the production of 233U (Figure 2). However, the parent nuclide of 212Bi is the β−-emitter 212Pb, having a T1/2 of 10.6 h. The chelator TCMC is used with 212Pb and functions as an in vivo nanogenerator for the α-particle emitter 212Bi. The University of Alabama (USA) has started a clinical trial to evaluate 212Pb-TCMC-trastuzumab toxicity levels and anti-tumor efficacy in patients with HER-2 positive cancers in the abdominal cavity (11).
Astatine-211 (211At) decays with a T1/2 of 7.2 h and emits an α-particle in both of the two possible decay routs to the stable nuclide 207Bi. Scintigraphy and standard dosimetry are possible due to the accompanying γ-radiation. The limited availability is currently a main obstacle for a wider use of this nuclide, as it can only be cyclotron produced (43). It has been used in clinical trials, locally administered in surgical resection cavities and i.p. as previously discussed (9, 10).
Dosimetry was originally developed for radiation protection (44) and diagnostic imaging (45), but is now also needed for optimization of the therapeutic situation using radiopharmaceuticals. The basic concepts of dosimetry are presented in two Medical Internal Radiation Dose (MIRD) publications (46, 47).
α-Particle dosimetry takes into account a number of different parameters, particularly the short path length of α-particles in tissue (~100 μm) and the inhomogeneous distribution of α-radiopharmaceuticals in tumors and tissues. Thus, predicting the biological effect based on mean absorbed dose in a tumor or organ might be misleading in some circumstances. The high-LET (~100 keV/μm) and varying LET (with a maximum at the Bragg peak) along the α-particle track are also parameters that have to be taken into account when performing α-particle dosimetry.
The RBE of α-particles ranges from 3 to 7, i.e., α-particle irradiation is 3–7 times more therapeutically effective, or toxic, per unit of absorbed dose than photons or electrons (47). In TAT clinical studies, an RBE of five has been applied to estimate the equivalent absorbed doses (10, 14, 48). The weighting factor applied when estimating the effective (or equivalent) absorbed dose (expressed in Sv, Sievert) is related to the stochastic effects of radiation, e.g., cancer induction. A factor of 20 is commonly recommended for the stochastic effects of α-particles that should however not be used when predicting the therapeutic efficacy or toxicity in patients who receive TAT treatment. Indeed, this weighting factor was conservatively derived for radiation protection and was never meant for estimating the deterministic effects relevant to therapy (47). Also, the clinical experience with α-particles is sparse, and therefore the tolerance to absorbed doses in humans has yet to be determined.
α-Particle dosimetry in the clinic require pharmacokinetic data similar to those that are required for conventional β–-particle therapies (22), e.g., urine, blood, and peritoneal fluids in the case of i.p. treatment (10). All α-particle emitters used so far in clinical studies (211At, 213Bi, 223Ra, 212Bi, and 225Ac) emit γ-photons, characteristic X-ray, or bremsstrahlung radiation. Using the γ-camera makes quantification of biodistribution possible. The spatial resolution of such images is, however, fairly low. Also, the injected activity is much lower than in a diagnostic setting, generally resulting in a poor signal-to-noise ratio. For similar reasons, 3-Dimensional single-photon emission computerized tomography (SPECT) imaging of the activity distribution in patients is time-consuming. The accuracy could be increased using co-registration techniques with computed tomography (CT) images (49).
Obviously, the absorbed dose in tumors and normal tissues need to be estimated from preclinical studies before initiating treatment studies. However, clinical quantification with the γ-camera can only give an estimate of the uptake of the radiopharmaceutical in whole organs and in macroscopic tumors, while quantification of the absorbed dose in smaller compartments in organs or microscopic tumors is hardly achievable. In TAT, the targeted tumors are often too small to be detected and, at best, indirect methods can be used for estimating the absorbed dose.
With regard to normal tissue protection, in certain cases, blocking agents can be used. For example, both astatine and iodine belong to the halogen elements and pre-treatment with potassium perchlorate can effectively prevent uptake of free 211At in cells expressing the sodium-iodine symporter (NIS), e.g., in the thyroid (10).
In the case of i.p. TAT for ovarian cancer, a control γ-camera image of the abdominal region with a radioactive-tracer analog to assure free distribution of the fluids is important. The radioactive flow out of the abdominal cavity can also be determined using a radioactive-tracer analog, by monitoring the activity concentration in blood over time (10). Pharmacokinetic data show that the variation in the absorbed dose in bone marrow can be around 20% (10). If the bone marrow is the dose-limiting organ, its absorbed dose then determines the maximal tolerated activity (MTA), and a radioactive-tracer analog study will be crucial for estimating the patient-specific MTA. However, for i.p. TAT, no effect on the hematopoiesis was recorded (10). Instead, other organs might determine the MTA, possibly the peritoneum; therefore, the activity concentration in the peritoneal fluid is crucial to calculate.
α-Particle dosimetry on the cell level should be used when macrodosimetry cannot explain the results of an experiment or when it adds value to the macrodosimetric method (50). For α-particles, the biological effect of just a single ionization event could be so large that the calculation of the mean absorbed dose in a tumor as a whole can be very misleading.
Hence, there is a need for microdosimetry when the statistical variation of the deposited radiation is not minimal in the target such as a cancer cell nucleus. The conceptual framework of microdosimetry that takes into account the stochastic nature of energy deposits in small microscopic targets was proposed almost 60 years ago (51), and the International Commission on Radiation Units and Measurements (ICRU) report No. 36 from 1983 defined all the microdosimetric concepts. Calculations and experiments have shown that as few as five high-LET α-particle traversals through the cell nucleus are enough to kill a cell, whereas 10,000–20,000 low-LET β–-particles are needed to achieve the same biological effect (52–54).
Importantly, microdosimetry should be considered for non-targeted but critical tissues, even if it receives a very low mean absorbed dose (47).
The way high-LET radiation like α-particles interact with biological matter has been described earlier (53, 55–60). They produce dense ionizations along a linear track and generate locally multiple damage sites in sensitive targets like DNA. These lesions, produced in close proximity to each other, are poorly repairable, thus making α-particles highly deleterious (61, 62). While conventional EBRT is characterized by high absorbed doses delivered in a very short time in a homogenous way, TAT and radionuclide therapy in general are characterized by a low absorbed dose rate, protracted exposure, and heterogeneous energy deposit (63).
In EBRT, physical events predominate in the final outcome of the therapy, and most of the effects can be correlated to the absorbed dose according to a linear, linear-quadratic, or sigmoid relationship. Conversely, physical characteristics of targeted radionuclide therapy can offer the cells the opportunity to repair some of their sublethal lesions (64–67). Nuclear DNA plays a central role in response to targeted radionuclide therapy, but other cellular sub-compartments including the mitochondria and cell membrane might also be strongly involved in situations of heterogeneous energy deposits (68–74). Therefore, the biology of the irradiated tissue and its interaction with its environment might play an even more pronounced role in targeted radionuclide therapy than EBRT, and bystander and abscopal effects involving activation of signaling pathways and the immune system should probably be investigated more accurately (75–77). The consequences are that the absorbed dose-effect might be more difficult to establish and radiation-induced biological effects might be observed in tissues far beyond the physical path length of the α-particles.
All targeted therapies rely on the ability of the vector to find its target and to allow the associated cytotoxic agent to deliver the cell-killing effect. Advances in genetic engineering have led to the development of many molecules that can be radiolabeled and used for RIT. However, despite the growing number of designed antibody fragments and fusion proteins, treatments are often hampered by less than optimal pharmacokinetics. The key lies in finding a balance between tumor radiation uptake and removal of circulating radioactivity. Rapid clearance of unbound radioimmunoconjugates is essential for limiting the absorbed dose to normal organs, but a too short a retention time in blood will result in a too short targeting time, and thus in the delivery of a too low absorbed dose to malignant cells.
This pharmacokinetic challenge can be handled by separating physically and temporally the targeting phase from the delivery of the ionizing radiation, an approach generally referred to as pre-targeted radioimmunotherapy (PRIT) (78, 79). A number of PRIT regimens, all based on the same essential principle, have been proposed since the pre-targeting concept was proposed by Goodwin et al. in 1988 (80). In the first step, a targeting immunoconjugate (pre-targeting molecule) is administered and sufficient time is allowed for its localization at tumor-associated antigen sites. As the pre-targeting molecule does not carry any cytotoxic substance, normal tissues are not affected by lengthy circulation times during the distribution phase. Then, unbound immunoconjugate molecules can be removed from the circulation using a clearing agent, before injecting the radiolabeled vector (effector molecule). The effector molecule is a small molecule designed to rapidly diffuse into tumors and cancer cell clusters, where it will specifically bind to the antigen-associated pre-targeting molecules. The fast clearance of unbound effector molecules improves the tumor-to-normal tissue ratios of absorbed dose compared with directly labeled immunoconjugates. With pre-targeting, no trade-off needs to be made between efficient targeting/penetration/tumor residence time and protection of dose-limiting normal tissues.
Efficient interaction between the pre-targeting molecule and the effector molecule has been achieved using a handful of techniques, particularly those based on streptavidin-biotin (81) or bispecific antibodies (82). Of the radionuclides with potential use in TAT, some appear more suitable than others when factors such as availability and daughter nuclides are taken into account, in addition to chelation and conjugation chemistry. In particular, two promising candidates for efficient therapy emerge: 211At and 213Bi. However, they both have short T1/2 (7.2 h and 45.6 min, respectively), which put high demands on the distribution of radiolabeled vectors to ensure favorable absorbed dose ratios. This issue could be overcome by using a pre-targeting strategy, thereby increasing the therapeutic potential of these short-lived α-particle emitters.
Several preclinical studies have shown the benefits of pre-targeted α therapy (PTAT), mainly in hematological cancers, such as AML (83), non-Hodgkin lymphoma (84), anaplastic large cell lymphoma (85), and adult T-cell leukemia (85). PTAT for disseminated ovarian carcinoma was evaluated in one study in which 211At-PRIT (1.5 MBq) and 211At-RIT (0.9 MBq) were compared in a mouse model of i.p. TAT (86). The administered activities were based on the previously estimated MTAs for the two regimens and resulted in equal tumor-free fractions (TFF; 0.45) 8 weeks after irradiation; however, the mice treated with 211At-PRIT had smaller tumors and lower ascites incidence. This indicates that pre-targeting can improve the outcome also of i.p. TAT, although the greatest gain of PTAT is generally considered to be in systemic treatments.
Radioimmunotherapy with short-ranged, high-efficiency α-particles is a very attractive and promising treatment strategy. α-Particles have an advantage in targeted therapy because of their exceptionally high cell-killing ability. Therefore, different from RIT with β–-particles, α-particle emitters labeled to a targeting vector can directly kill single cancer cells (by self-irradiation). Several completed or on-going clinical trials using TAT have shown its feasibility for treating disseminated and/or micro-metastatic malignancies without significant or insurmountable problems of toxicity. Although the definition of micrometastases is vague, in clinical oncology occult metastases (i.e., not detected by routinely used imaging procedures) might involve single tumor cells up to clusters of billions of cells. Therefore, a cocktail of both α- and β–-emitting radioconjugates might be more effective in some cases.
The possibility of TAT as a potential curative treatment includes its use as a local boost after initial treatment (e.g., i.p. in EOC), or perhaps as i.v. systemic adjuvant treatment, both targeting micro-metastatic disease. A systemic approach may indeed be of particular interest in patients with EOC that includes retroperitoneal vascularized metastases, e.g., in the lymph nodes. Fractionated RIT is another potentially interesting regimen to improve the therapeutic index, thus resulting in reduced normal organ toxicity while maintaining the therapeutic efficacy (87). Radionuclides that emit Auger electrons could offer an alternative approach compared with the nuclides described in this article, reviewed elsewhere (88). Auger electrons are energetically very weak (< <1 keV) and have a path length in tissue that is far shorter than that of α-particles. However, to effectively damage DNA molecules, the Auger emitter has to bind to the DNA.
The therapeutic outcome of TAT is influenced by a number of crucial issues that all need to be handled, e.g., the specificity of the antibody/targeting construct; the level of antigenic expression on the tumor cells; the potential loss of immunoreactivity of the antibody/targeting construct; the amount of unlabeled antibody/targeting construct after injection; the existence of diffusion barriers that hinder the penetration of the antibody/targeting construct into the tumors; the choice of radionuclide (half-life and path length); too low specific radioactivity; and for the i.p. situation, any extra peritoneal location of tumor cells.
A major issue that may hamper wide implementation in the clinic and that needs to be simultaneously addressed is the availability of suitable α-particle emitters at a reasonable cost (43, 89). Otherwise, TAT will remain just a potentially effective treatment, or a very rarely implemented option. Finally, after safety issues and pharmacokinetics have been established, for all types of malignancies that might benefit from TAT/PTAT, we need to conduct randomized, controlled, clinical studies. These need to include a high enough number of patients to allow meaningful comparison and evaluation of different treatment strategies.
The authors declare that the research was conducted in the absence of any commercial or financial relationships that could be construed as a potential conflict of interest.
The information comes from:https://www.ncbi.nlm.nih.gov/pmc/articles/PMC3890691/