Measurement of Fission Neutron Spectrum Averaged Gross Sections of some Threshold Reactions on Ruthenium: Small Scale Production of 96Tn in a Nuclear Reactor

Nuclear Sample Provenance and Age Determination Using Ruthenium Isotopes

Measurements of the ruthenium isotopic composition of nuclear samples could provide information about the method of sample production, sample irradiation history, and age. To investigate the feasibility and applicability of this idea, this study focuses on measurements of the ruthenium isotope composition of a nominally single-isotope ¹⁰⁶Ru radioactivity standard, where the complications of environmental mixing are eliminated. The measurements of the ¹⁰⁶Ru standards reveal unusual stable ruthenium isotopic compositions consistent with fissiogenic ruthenium. Three different lots of the material have been investigated, and the isotopic composition is found to be different for lot 1 as compared to lots 2 and 3, indicating a longer irradiation duration incurred during the production of lot 1. Through measurements of ¹⁰⁶Ru and its ¹⁰⁶Pd daughter, radiochronometry can be used to infer the ages of the samples. Lot 1 is older than lots 2 and 3 and was produced 4.91(5) years before the reference date of 1/1/21, approximately 2.7 years before lots 2 and 3. In an effort to better understand the sample production pathway, the isotopic measurements are compared with nuclear reactor simulations, which suggest that the material was generated by irradiation of a low-enriched uranium target material in a light water reactor. These findings have significant implications for nuclear treaty monitoring, providing an example of the power of ruthenium isotope measurements to discern details of sample origin and history.

95gTc and 96gTc as alternatives to medical radioisotope 99mTc

We studied 95gTc and 96gTc as alternatives to the medical radioisotope 99mTc. 96gTc (95gTc) can be produced by (p, n) reactions on an enriched 96Mo (95Mo) target with a proton beam provided by a compact accelerator such as a medical cyclotron that generate radioisotopes for positron emission tomography (PET). The γ-rays are measured with an electron-tracking Compton camera (ETCC). We calculated the relative intensities of the γ-rays from 95gTc and 96gTc. The calculated γ-ray intensity of a 96gTc (95gTc) nucleus is as high as 63% (70%) of that of a 99mTc nucleus. We also calculated the patient radiation doses of 95gTc and 96gTc, which were larger than that of 99mTc by a factor of 2-3 based on the applied assumptions. A medical PET cyclotron which can provide proton beams with energies of 11-12 MeV and a current of 100 μA can produce 12 GBq (39 GBq) of 96gTc (95gTc) for operation time of 8 h, which can be used for 240 (200) diagnostic scans.

Development of radiochemistry in Pakistan – 1960 to 2010

With the inception of Pakistan Atomic Energy Commission (PAEC) in 1956, peaceful uses of atomic energy commenced for the benefit of scientific community as well as masses of Pakistan. Radiochemistry played a vital role right from the beginning. The research and development in this field accelerated soon after the criticality of the first research reactor named as Pakistan Research Reactor (PARR- 1) at the Pakistan Institute of Nuclear Science and Technology (PINSTECH), Islamabad. The first radioisotope produced at PARR-1 for application in nuclear medicine was 131I. Later on, many other radioisotopes were prepared and radiopharmaceuticals were synthesised for their use in industry and hospitals. Besides providing pure radioactive tracers for nuclear medicine, radiochemistry also enhanced the detection limit of impurities at all stages of nuclear fuel cycle for power generation. In 1983, research in the field of nuclear data measurement began. The main aim was to identify suitable conditions for the production of radionuclides for cancer diagnostics, treatment and therapy. With the establishment of a second research reactor (PARR-2) at PINSTECH, research in neutron activation analysis, radioisotope production and separation studies gained more momentum and many research articles were published. Solvent extraction, adsorption and ion-exchange were the main routes of separation in those studies. Separation of heavy metals and treatment of waste generated in a nuclear power plant are other important aspects related to environmental restoration and nuclear waste management, where radiochemistry is required. In future, work in radiochemistry will be continued on similar lines to develop novel radiopharmaceuticals, identify indigenous schemes for nuclear waste management and work out intelligent procedures for material characterization for benefit to mankind, especially the people of Pakistan.

Fission spectrum averaged cross section measurements of some neutron threshold reactions of relevance to medical radionuclide production

A resume is given of radiochemical measurements of nuclear reaction cross sections relevant to the production of some medically important radionuclides carried out at PINSTECH during the last decade. Systematic studies on fission neutron spectrum averaged cross sections of several threshold reactions, like (n, 2n), (n, p) and (n, α), on titanium, ruthenium, europium and dysprosium for the production of 45Ti, 96Tc, 153Sm and 153Gd are described.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on dysprosium: small-scale production of no-carrier-added153Gd in a nuclear reactor

Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, 2n), (n,p) and (n, α) reactions on isotopes of dysprosium; our measurements constitute the first systematic studies. Of special interest was the investigation of the156Dy(n, α)153Gd reaction for the production of no-carrier-added153Gd in a nuclear reactor. Using 100% enriched156Dy target, 8633 KBq of153Gd per batch can be produced which is, however, not sufficient for medical application.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on europium: small scale production of no-carrier-added153Sm in a nuclear reactor

Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, 2n), (n, p) and (n, α) reactions on isotopes of europium. Our measurements constitute the first systematic studies. Of special interest was the investigation of153Eu(n, p)153Sm reaction for the production of no-carrier-added153Sm in a nuclear reactor. Using 100% enriched153Eu target, 97.21 MBq153Sm per batch can be produced which is, however, not sufficient for medical application.

Fission neutron spectrum averaged cross sections of some threshold reactions on cadmium: production feasibility of no-carrier-added 103Pd in a nuclear reactor

Systematic studies on fission neutron spectrum averaged cross sections of some threshold reactions like (n,p) and (n,α) on cadmium were carried out using the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy. Special attention was paid to the formation of 103Pd via the 106Cd(n,α) 103Pd reaction since it is an important therapeutic radionuclide. At a fast flux neutron density of 7.5 × 1013 cm2s-1 and an irradiation time of 120 h, using 100% enriched 106Cd target 340 MBq of no-carrier-added 103Pd per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide.

Measurement of fission neutron spectrum averaged cross sections of some threshold reactions on zirconium: Production possibility of no-carrier-added 90Y in a nuclear reactor

Employing the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy fission neutron spectrum averaged cross sections were measured for several (n, p) and (n, α) reactions on isotopes of zirconium; our measurements constitute the first systematic study. The special interest was in the investigation of the 90Zr(n, p)90Y reaction for the production of no-carrier-added 90Y in a nuclear reactor. At a fast neutron flux density of 7.5 × 1013 cm−2 s−1 and an irradiation time of 120 h, using 100% enriched 90Zr target 0.548 GBq of no-carrier-added 90Y per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide.

Fission neutron spectrum averaged cross sections of some threshold reactions on titanium: production feasibility of 45Ti in a nuclear reactor

Systematic studies on fission neutron spectrum averaged cross sections of some threshold reactions like (n, 2n), (n, p) and (n, α) on titanium were carried out using the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy. Special attention was paid to the formation 45Ti via the 46Ti(n, 2n)45Ti reaction since 45Ti labeled compounds are potentially important radiopharmaceuticals. At a fast flux neutron density of 7.5 × 1013 cm−2 s−1 and using 100% enriched 46Ti target, 8.58 MBq saturation activity of 45Ti per batch can be produced. Due to rather low activity and low specific activity the procedure is not feasible from medical application point of view.