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Electrodeposition of silver on copper substrate was carried out using alkaline plating baths for production of Cadmium-109 via the nat Ag( p, n) 109 Cd nuclear reaction. The target could withstand 15 MeV proton beam currents of up to 150 µA; the 109 Cd production yield was 1.88 µCi/µA h (0.069 MBq/µA h). Cadmium-109 was separated from silver and non-isotopic impurities by ion-exchange chromatography.

Gallium-68 (T 1/2 = 68 min, I β+ = 89%) is an important positron-emitting radionuclide for positron emission tomography and used in nuclear medicine for diagnosing tumours. This study gives a suitable reaction to produce 68 Ga. Gallium-68 excitation function via 68 Zn(p, n) 68 Ga, 68 Zn(d, 2n) 68 Ga, 70 Zn(p, 3n) 68 Ga and 65 Cu(α, n) 68 Ga reactions were calculated by ALICE-91 and TALYS-1.0 codes. The calculated excitation function of 68 Zn(p, n) 68 Ga reaction was compared with the reported measurement and evaluations. Requisite thickness of the targets was obtained by SRIM code for each reaction. The 68 Ga production yield was evaluated using excitation function and stopping power.

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Decrease of the economically accessible uranium resources and the inherent proliferation resistance of thorium fuel motivate its application in nuclear power systems. Estimation of the nuclear reactor’s neutronic parameters during different operational situations is of key importance for the safe operation of nuclear reactors. In the present research, thorium oxide fuel burn-up calculations for a demonstrative model of a heavy water- -cooled reactor have been performed using MCNPX 2.6 code. Neutronic parameters for three different thorium fuel matrices loaded separately in the modelled thermal core have been investigated. 233U, 235U and 239Pu isotopes have been used as fissile element in the thorium oxide fuel, separately. Burn-up of three different fuels has been calculated at 1 MW constant power. 135X and 149Sm concentration variations have been studied in the modelled core during 165 days burn-up. Burn-up of thorium oxide enriched with 233U resulted in the least 149Sm and 135Xe productions and net fissile production of 233U after 165 days. The negative fuel, coolant and void reactivity of the used fuel assures safe operation of the modelled thermal core containing (233U-Th) O2 matrix. Furthermore, utilisation of thorium breeder fuel demonstrates several advantages, such as good neutronic economy, 233U production and less production of long-lived α emitter high radiotoxic wastes in biological internal exposure point of view

The accelerator driven system is an innovative reactor which is being considered as a dedicated high-level waste burner. The function of the spallation target in accelerator driven system is to convert the incident high-energy particle beam to low-energy neutrons. One of the quantities of most interest for practical purposes is the number of neutrons produced per proton in a spallation target. However, this vital value depends not only on the material, but on the size of the target as well, due to the internuclear cascade. The MCNPX 2.4 code can be used for spallation target computation. Some benchmark results have been compared with MCNPX 2.4 simulations to verify the code's potential for calculating various parameters of an accelerator driven system target. Using the computation method, neutron interaction processes such as loss, capture and (n, xn) into a spallation target have been studied for W, Ta, Pb, Bi, and LBE spallation targets in different target dimensions. With relative errors less than 10%, the numerical simulation provided by the MCNPX code agrees qualitatively with other simulation results previously carried out, qualifying it for spallation calculations. Among the studied targets, W and Ta targets resulted in a higher neutron spallation yield using lesser target dimensions. Pb, Bi, and LBE spallation targets behave similarly regarding the accessible leaked neutron yield on the outer surface of the spallation target. By use of a thicker target, LBE can compete with both W and Ta targets regarding the neutron yield parameter

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