Simulation and analysis of the medical radioisotope 177Lu production based on a high-intensity DT neutron generator

177Lu is one of the medical radioisotopes with wide application prospects. To explore the 177Lu production besides using reactors, a concept of utilizing high-enriched cylindrical 176Lu based on a high-intensity DT neutron generator via the direct neutron capture reaction has been proposed. The production of 177Lu in a176Lu2O3 solid shell target has been simulated considering layers of neutron multiplier, moderator, and reflector around itself. Materials used as multipliers, moderators, and reflectors have been investigated and analyzed for neutronics optimization. The combinations and thicknesses of different layers have been calculated using the response surface methodology (RSM), and the three layers were found almost independently for the 177Lu production. Both the activity and the specific activity of produced 177Lu were improved significantly using the RSM. According to the calculations, about 13.5 Ci 177Lu can be produced after 40 days of irradiation with a stable neutron yield of 5 × 1011 n/s.

Preliminary experiments to produce lutetium-177 in the TRR-1/M1 Thai research reactor

Lutetium-177 has emerged as a highly efficient radionuclide for medical applications, particularly in the field of targeted radionuclide therapy. Its production has been increasingly optimized through neutron activation techniques, which offer distinct advantages over alternative methods. Utilizing the TRR-1/M1 research reactor, which has been in operation for nearly six decades, provides a strategic opportunity for advancing domestic radioisotope production, thereby supporting the medical sector in Thailand. The TRR-1/M1 reactor, despite its operational age, continues to exhibit considerable potential for contributing to medical research and radioisotope development in Thailand. Preliminary experimental results, conducted at a flux of 1.42 × 1012 n/cm2/s demonstrated promising outcomes, even under operational constraints such as fuel management limitations. Notably, the direct neutron activation of natural lutetium oxide notably yielded a specific activity of 177Lu at 10.92 GBq/g (295.06 mCi/g) with a production yield of 44.8%, with projections reaching 222 GBq/g (6 Ci/g) after 40 days of neutron irradiation. In comparison, the indirect method, using natural ytterbium oxide as a precursor, achieved a maximum specific activity of 177Lu at 6.6 MBq/g (180.3 μCi/g) with a yield of 37.8% of a theoretical maximum of 17.6 MBq/g (476 μCi/g) after only 10 h of neutron activation. These results highlight the feasibility and promise of 177Lu radioisotope production in Thailand.

Need for enrichment of lutetium isotope and design of a laser based separator module

Lutetium-177 radio-pharmaceutical has become an important theranostic candidate in cancer treatment. Its availability from bench-to-bed requires strategic implementation of isotope-enrichment, neutron-irradiation and radio-chemical techniques. In this paper, the need for enrichment of lutetium-176 is emphasized by estimating specific activity of lutetium-177 as a function of enrichment percentage for typical neutron flux available at Dhruva reactor, India. A novel Atomic Vapour Laser Isotope Separation (AVLIS) module for lutetium-176 enrichment is designed to meet the above requirement. The paper documents its characteristics and production estimates. The design is carried out after critical assessment and evaluation of available AVLIS-infrastructure in the country. Outline of lutetium-177 enrichment, capable of producing non-carrier-added lutetium is also provided. This work concludes that India has taken a step forward towards self-reliance (Atmanirbhar Bharat) in securing the supply chain of lutetium-177.

Preparation of Patient Doses of [177Lu]Lu-DOTATATE and [177Lu]Lu-PSMA-617 with Carrier Added (CA) and No Carrier Added (NCA) 177Lu

Purpose

[177Lu]Lu-DOTATATE and [177Lu]Lu-PSMA-617 used for targeted radionuclide therapy are very often prepared in the hospital radiopharmacy. The preparation parameters vary depending upon the specific activity of the 177Lu used. The aim of this study was to develop optimized protocols to be used in the nuclear medicine department for the preparation of patient doses of the above radiopharmaceuticals.

Method

177Lu (CA and NCA) were used for radiolabeling DOTATATE and PSMA-617. Parameters studied are 177Lu of different specific activity and different peptide concentrations and two different buffer systems. Paper and thin layer chromatography systems were used for estimating the radiochemical yield as well as radiochemical purity. Solid-phase extraction was used for the purification of the labeled tracers.

Results

[177Lu]Lu-DOTATATE was prepared with CA 177Lu (n = 13) and NCA177Lu (n = 6). Four batches each of [177Lu]Lu-PSMA-617 were prepared using CA and NCA 177Lu. Radiochemical yields > 80% and final product with less than < 1% radiochemical impurity could be obtained in all batches which were used for therapy.

Conclusion

Robust protocols for the preparation of clinical doses of [177Lu]Lu-DOTATATE and [177Lu]Lu-PSMA-617 were developed and used for the preparation of clinical doses. The quality of the SPECT images of both the tracers are consistent with the expected uptake in respective diseases.

Theragnostic radionuclides: a clinical perspective

The concept of theragnostics goes back to the earliest days of nuclear medicine, with [¹²³I/¹³¹I]iodide in thyroid disease and [¹²³I/¹³¹I]MIBG in phaeochromocytoma being examples in long-term use. However, in recent years there has been a great expansion in the application of theragnostics, beginning with [⁶⁸Ga/¹⁷⁷Lu]-labelled somatostatin peptides for evaluation and treatment of neuroendocrine tumors. We are currently seeing the rapid development of [⁶⁸Ga/¹⁷⁷Lu]PSMA theragnostics in metastatic prostate cancer. While these applications are very promising, there are a number of practicalities which must be addressed in the development and introduction of novel theragnostics. The physical half-lives of the diagnostic and therapeutic radionuclides must be appropriate for imaging and delivery of targeted cell killing, respectively. The types of radioactive emissions are critical; beta particles can traverse several millimeters but also risk damaging non-target tissues, while alpha particles deliver their energy over a much shorter path length, a few cell diameters, and must be more directly targeted. It must be practical to produce the therapeutic radionuclide and the final radiopharmaceutical and deliver them to the final user within an appropriate time-frame determined by half-life and stability. The biodistribution of the agent must demonstrate adequate accumulation and retention in the target tissue with clearance from adjacent and/or radio-sensitive normal tissues. The commercial success of recently introduced theragnostics suggests a rosy future for personalized medicine.

Analysis of lutetium-177 production at the WWR-K research reactor

Production of lutetium-177 using direct nuclear reaction ¹⁷⁶Lu(n,γ)¹⁷⁷Lu by WWR-K reactor neutrons on enriched LuCl₃ (up to 82% of ¹⁷⁶Lu) is described. Calculations were performed by MCNP6 transport code. Two different irradiation positions of the WWR-K research reactor were considered. Estimates of the maximum specific activity of the luthetium-177 are obtained for the reactor irradiation positions located: (a) in the reactor core centre, (b) in the core periphery. In these positions, thermal neutron flux is two times different. Experimental data was shown that k-factor is 1.5 for considered irradiation positions. The study shows that for the position located in the core center, the estimated maximum specific activity of lutetium-177 is 819 GBq/mg, is to be achieved after 15 days of irradiation. For the position located in the core periphery, specific activity of lutetium-177 is 561 GBq/mg, is to be achieved after 20 days of irradiation. Ratio of Lu-177m to Lu-177 specific activity is not more than 0.025 for both irradiation positions.

A Step-by-Step Guide for the Novel Radiometal Production for Medical Applications: Case Studies with 68Ga, 44Sc, 177Lu and 161Tb

The production of novel radionuclides is the first step towards the development of new effective radiopharmaceuticals, and the quality thereof directly affects the preclinical and clinical phases. In this review, novel radiometal production for medical applications is briefly elucidated. The production status of the imaging nuclide 44Sc and the therapeutic β--emitter nuclide 161Tb are compared to their more established counterparts, 68Ga and 177Lu according to their targetry, irradiation process, radiochemistry, and quality control aspects. The detailed discussion of these significant issues will help towards the future introduction of these promising radionuclides into drug manufacture for clinical application under Good Manufacturing Practice (GMP).