Use of cyclotrons in medicine

Performance evaluation of Gallium-68 radiopharmaceuticals production using liquid target PETtrace 800 cyclotron

Due to increased demand, cyclotron has an expanding role in producing Gallium-68 (68Ga) radiopharmaceuticals using solid and liquid targets. Though the liquid target produces lower end-of-bombardment activity compared to the solid target, our study presents the performance of 68Ga radiopharmaceuticals production using the liquid target by evaluating the end-of-bombardment activity and the end-of-purification activity of [68Ga]GaCl3. We also present the effect of increasing irradiation time, which significantly improves the end-of-synthesis yield. From the result obtained, the end-of-bombardment activity produced was 4.48 GBq, and the [68Ga]GaCl3 end-of-purification activity produced was 2.51 GBq with below-limit metallic impurities. Increasing the irradiation time showed a significant increase in the end-of-synthesis activity from 1.33 GBq to 1.95 GBq for [68Ga]Ga-PSMA-11 and from 1.13 GBq to 1.74 GBq for [68Ga]Ga-DOTA-TATE. Based on the improvements made, the liquid target production of 68Ga radiopharmaceuticals is feasible and reproducible to accommodate up to 5 patients per production. In addition, this work also discusses the issues encountered, together with the possible corrective and preventative measures.

Status of the decay data for medical radionuclides: existing and potential diagnostic γ emitters, diagnostic β+ emitters and therapeutic radioisotopes

Recommended half-lives and specific well-defined emission energies and absolute emission probabilities are important input parameters that should be well-defined to assist in ensuring the diagnostic and therapeutic efficacy of individual radionuclides when applied in the field of nuclear medicine. Bearing in mind the nature of these requirements, approximately one hundred radionuclides have been considered and re-assessed as to whether their decay data are either adequately quantified, or require further in-depth measurements to improve their existing status and merit full re-evaluations of their decay schemes. The primary aim of such a review is to provide sufficient information on the existing and future requirements for such atomic and nuclear data.

An overview of nuclear data standardisation work for accelerator-based production of medical radionuclides in Pakistan

The standardisation of nuclear reaction cross section data is an integral part of optimisation of production routes of medical radionuclides. The production cross sections are available for the reactor and cyclotron produced radionuclides to be used for diagnostics or therapeutic procedures. The types of nuclear data needed, and the sources of their availability are summarized. The method of standardisation of charged-particle data is briefly described. A historical overview of research work in Pakistan in this direction is given. Examples of a few medically important radionuclides, such as 64Cu, 86Y, 89Zr, 103Pd, 186Re, etc., whose data were standardised and evaluated are highlighted. Calculated thick target yields from the recommended data are given. Some new directions in the nuclear data research are outlined.

Flux-weighted average cross sections of medical isotopes in the 127I(γ,xn) reactions with the bremsstrahlung end-point energies of 50 and 70MeV

The flux-weighted average cross sections of ¹²⁷I(γ,xn)^126,124,123I reactions were experimentally determined with the bremsstrahlung end-point energies of 50 and 70 MeV by activation and off-line γ-ray spectrometric technique, using the 100 MeV electron linac at the Pohang Accelerator Laboratory, Korea. The ¹²⁷I(γ,xn)^126-121I reaction cross sections as a function of photon energy were theoretically calculated using the TALYS 1.8 code. The flux-weighted average cross sections of ¹²⁷I(γ, xn)^126,124,123I reactions as a function of bremsstrahlung end-point energies were also obtained from the literature and TALYS data based on mono-energetic photons and are found to be in close agreement with the present data. The experimental and theoretical ¹²⁷I(γ, xn)^126-121I reaction cross-sections increase with bremsstrahlung energy, which indicates the role of excitation energy. After a certain energy, the individual ¹²⁷I(γ, xn)^126-121I reaction cross sections decrease with increase of bremsstrahlung energy due to opening of other reactions, which indicates sharing of energy in different reaction channels. The radioactive iodine isotopes production cross-sections are important for the useful medical purpose.

Evaluation of nuclear reaction cross sections for optimization of production of the emerging diagnostic radionuclide ⁵⁵Co

The excitation functions of the (54)Fe(d,n)(55)Co, (56)Fe(p,2n)(55)Co and (58)Ni(p,α)(55)Co reactions were analyzed with relevance to the production of the β(+)-emitter (55)Co (T½=17.53 h), a promising cobalt radionuclide for PET imaging. The nuclear model codes ALICE-IPPE, EMPIRE and TALYS were used to check the consistency of the experimental data. The statistically fitted excitation function was employed to calculate the integral yield of the product. The amounts of the radioactive impurities (56)Co and (57)Co were assessed. A comparison of the three investigated production routes is given.

Synthesis, characterization and in vivo evaluation of [(62)Zn]-benzo-δ-sultam complex as a possible pet imaging agent

OBJECTIVE

The development of a new tracer based on the cyclic sulfonamides (sultams) was investigated.

METHODS

3-(Methoxy-phenyl-methyl)-1,6-dimethyl-1H benzo[c][1,2] thiazine 2,2-dioxide (benzo-δ-sultam) was synthesized and characterized by elemental analysis, FT-IR spectroscopy and single crystal X-ray structure determination. The prepared cyclic sulfonamide was labeled with non-commercial (62)Zn radioisotope for fast in vivo targeting and Coincidence imaging purposes (radiochemical purity 97 % ITLC, 96 % HPLC, specific activity 20-23 GBq/mmol). In vivo biodistribution of the final complex was investigated in Sprague Dawley(®) rats bearing fibro sarcoma tumor after 2, 4 and 8 h post injection and compared with free Zn(+2) cation.

RESULTS

Using instant paper chromatography method, the physicochemical properties of labeled compounds were found sufficiently stable in organic phases, e.g. a human serum, to be reliably used in bioapplications.

CONCLUSIONS

The complex exhibited a rapid as well as high tumor uptake (tumor to blood ratio 4.38 and tumor to muscle ratio 9.63) resulting in an efficient tumor targeting agent.

Distribution of thermal neutron flux around a PET cyclotron

The number of positron emission tomography (PET) examinations has greatly increased world-wide. Since positron emission nuclides for the PET examinations have short half-lives, they are mainly produced using on-site cyclotrons. During the production of the nuclides, significant quantities of neutrons are generated from the cyclotrons. Neutrons have potential to activate the materials around the cyclotrons and cause exposure to the staff. To investigate quantities and distribution of the thermal neutrons, thermal neutron fluxes were measured around a PET cyclotron in a laboratory associating with a hospital. The cyclotron accelerates protons up to 18 MeV, and the mean particle current is 20 μA. The neutron fluxes were measured during both 18F production and C production. Gold foils and thermoluminescent dosimeter (TLD) badges were used to measure the neutron fluxes. The neutron fluxes in the target box averaged 9.3 × 10(6) cm(-2) s(-1) and 1.7 × 10(6) cm(-2) s(-1) during 18F and 11C production, respectively. Those in the cyclotron room averaged 4.1 × 10(5) cm(-2) s(-1) and 1.2 × 10(5) cm(-2) s(-1), respectively. Those outside the concrete wall shielding were estimated as being equal to or less than ∼3 cm s, which corresponded to 0.1 μSv h(-1) in effective dose. The neutron fluxes outside the concrete shielding were confirmed to be quite low compared to the legal limit.

Evaluation of charged particle induced reaction cross section data for production of the important therapeutic radionuclide 186Re

186Re is an important radionuclide having a half-life of 3.72 d that is suitable for radioimmunotherapy. Its production in no-carrier-added form is done via charged particle induced reactions and the data are available in EXFOR library. We evaluated two charged particle induced reactions, namely 186W(p,n) 186Re and 186W(d,2n) 186Re. In the first case, analysis was done up to about 70 MeV but in the latter only up to about 50 MeV. A statistical procedure supported by nuclear model calculations using the codes STAPRE, EMPIRE and TALYS was used to validate and fit the data. The recommended sets of data derived together with 95% confidence limits are reported. The application of those data, particularly in the calculation of integral yields is discussed. The 186W(p,n) 186Re reaction on highly enriched 186W is presently the method of choice for production of no-carrier-added 186Re and, taking into account the radionuclidic purity, the maximum recommended proton energy is 18 MeV. The formation of the very long-lived isomer, 186mRe, is briefly discussed. The 186W(d,2n) 186Re reaction could also be interesting if a high-intensity accelerator would be available.

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

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

Neutron dose rate in the facility at the cyclotron center of Chung Shan Medical University

The neutron dose equivalent rate (DR) leaking from the self-shielded cyclotron was measured using an FHT-751 neutron counting system in the facility at the Cyclotron Center of Chung Shan Medical University. This system was calibrated using two (252)Cf neutron sources and simulated according to MCNP code. The results show various DRs up to 120 microSv h(-1) in the cyclotron room. Two-dimensional distributions of measured neutron DRs indicate an explicit, heavy leakage of neutrons through the self-shielded interlock. The neutron DR of the operating cyclotron that is hazardous to the health of medical personal and the public is evaluated.