Potential production of Ac-225 at Egyptian second research reactor (ETRR-2) through neutron induced transmutation of Ra-226

A considerable focus has been paid to the production of 225Ac due to its effective therapeutic action in alpha-targeted radiotherapy. Considering the future global clinical demand, it is necessary to increase the production capacity of 225Ac. A feasibility study was conducted to investigate the production of 225Ac through neutron induced transmutation of 226Ra at the Egyptian Second Research Reactor (ETRR-2) using the MCNPX code. The calculations were carried out for 1 g of 226Ra target exposed to the highest neutron flux in the irradiation grid surrounding the reactor core. The 227Ra, 225Ra, 227Ac, and 225Ac generated activities as a function of irradiation and decay times were estimated. Our study revealed that in this non-linear production process, 39.22 MBq of pure 225Ac could be obtained after three days of irradiation, while 148.74 MBq could be obtained after fifteen days of continuous irradiation.

Feasibility study on production of Sc-47 from neutron irradiated Ca target for cancer theranostics applications

Scandium-47 is one of the most useful radioisotopes which is gaining great importance in cancer theranostics applications due to its favorable nuclear and chemical properties. MCNPX2.7.0 code was used to simulate the neutron activation of natural calcium target positioned at a thermal neutron flux of 1.8 × 1014 n cm−2 s−1 in the Egypt Second Research Reactor (ETRR-2). The burn card was used to calculate 47Ca and 47Sc radioactivities during 3 days irradiation and 20 days post-irradiation. The undesirable impurities generated during this period were also calculated. The obtained calculations were found to be in agreement with the experimental measurements. The distribution coefficient value (Kd) of 47Sc(III) as well as 47Ca(II) ions was determined using the commercially available ion-exchanger Chelex 100 in HNO3 and/or HCl media. Radiochemical separation of 47Sc(III) from 47Ca(II) was studied using HNO3 and HCl solutions and the results showed that HNO3 is a better medium than HCl for complete retention and recovery of 47Sc(III), where the recovery yields were 85 ± 1.2 and 95 ± 0.87 % using 1 M HCl and 1 M HNO3 solutions, respectively. The recovery yield obtained in our work was higher than in the reported procedures. Radionuclidic, radiochemical and chemical purities were investigated to ensure the suitability of 47Sc(III) for nuclear medicine applications.

Thermal hydraulic assessment of a new design of PWR fuel assembly

The recycling of the spent fuel of nuclear reactors is the important aspects in the fuel cycle technology. One of the recycling of the spent fuel is reuse of spent PWR (pressurized water reactor) fuel in CANDU reactor after re-fabricated into CANDU (CANda deuterium uranium) fuel bundles; this technology called Direct Use of Spent PWR Fuel In CANDU Reactor (DUPIC). This technology was modified however; it introduced a new design for the PWR fuel assembly. The new fuel can be transferred directly to CANDU reactors without processing, only after a certain cooling time. In this study the evaluation from thermal hydraulic point of view is done. A Westinghouse PWR fuel assembly was selected as a reference. Thermal hydraulic simulation for reference and new design fuel occurred by using RELAP5 (Reactor Excursion and Leak Analysis Program) code, then comparing the both results. The hottest rod was considered a worst condition in the reactor so the model simulated it rather than whole core. The new design achieves a good results comparing with a reference one.

Estimation of 99Mo production rates from natural molybdenum in research reactors

Molybdenum-99 is one of the most important radionuclides for medical diagnostics. In 2015, the International Atomic Energy Agency organized a round-robin exercise where the participants measured and calculated specific saturation activities achievable for the ⁹⁸Mo(n,γ)⁹⁹Mo reaction. This reaction is of interest as a means to locally, and on a small scale, produce ⁹⁹Mo from natural molybdenum. The current paper summarises a set of experimental results and reviews the methodology for calculating the corresponding saturation activities. Activation by epithermal neutrons and also epithermal neutron self-shielding are found to be of high importance in this case.