International Journal of Nuclear Energy Science and Technology (6 papers in press)
Evaluation of iodine-129 transmutation fraction in high flux reactors
by Fahim Tighemine, Naima Amrani, Ahmed Boucenna, A. Abdelghafar Galahom
Abstract: This study aims to transmute iodine-129 (I-129) into a stable isotope of xenon-130 (Xe-130) by irradiation in the high flux reactor. I-129 is considered one of the most hazardous long-lived fission products. The numerical results of the transmutation fraction, the transmutation amount of I-129 and mass evolution of Xe-130 produced in three high flux reactors, namely Petten, BR2 and SM3, are simulated for an irradiation time of 300 effective full power days using the ChainSolver 2.34 code. Based on the results obtained, it is recommended to use SM3 as the most effective I-129 transmutation application.
Keywords: transmutation fraction; long-lived fission product; iodine-129; iodine-127; ChainSolver 2.34; high flux reactor.
Production cross-sections calculations of proton and neutron induced reactions for producing copper radionuclide used in PET imaging and radiation therapy applications
by Nabeel F. Lattoofi, Ali A. Al Zubaidy
Abstract: In the last decades, many radioisotopes have been widely used in medical imaging diagnostic and radiation therapy applications. Among these are the copper isotopes, which are the most important isotopes used in these fields. These isotopes have been produced in reactors with different nuclear reactions in which a non-copper element has been used as a target. In our present research, we aimed to calculate the nuclear (p, n) and (n, p) reactions cross-section used in producing different copper isotopes, namely 60, 61, 62, 63, 64, 67Cu, using the Empire code for this purpose. The calculated results, investigated with different phenomenological and microscopic models for level density, have been compared with the available experimental data taken from the EXFOR database for energy ranging from 0 MeV up to 30 MeV. The results show a good agreement with experimental ones for (p, n) reactions with Hartree Fock Bogoliubov microscopic model (HFBM). However the cross section calculations with Enhanced Generalized Superfluid Model (EGSM) for (n ,p) reactions are closer to the experimental data.
Keywords: level density model; cross-section; therapeutic radioisotope.
Comparison of TRIGA reactor steady-state thermal-hydraulic predictions by COMSOL multiphysics with experimental data
by Ahmed K. Alkaabi, Jeffrey King
Abstract: This paper presents United States Geological Survey TRIGA one- and multiple-channel thermal hydraulic (TH) models developed using the COMSOL code to examine the effects of coolant cross-flow on coolant, cladding, and fuel temperatures. There are considerable variations in the profiles of the coolant axial temperatures and outlet temperatures as predicted by multiple-channel model from those predicted by the one-channel model. The one-channel model forecasts that the temperature of the coolant within the fuel rings increases axially with the height of the core, whereas the temperature of the coolant predicted by the multiple-channel model increases as a function of core height in the B-, C-, and D-rings, peaks and then reduces within the E-, F-, and G-rings. Within the multiple-channel model, the coolant appears to flow from the cores outermost opening at the lower side to the centre of the core. Finally, predictions of all models are benchmarked with the experimental data.
Keywords: TRIGA reactors; multiple-channel models; thermal hydraulic analysis.
Production calculations of 177Lu as a no-carrier-added radionuclide in Tehran Research Reactor
by Nafise Salek, Sara Vosoughi, Ali Bahrami Samani
Abstract: Owing to the suitable nuclear decay characteristics, 177Lu is an attractive radionuclide for various therapeutic applications. The non-carrier-added (NCA) form of 177Lu has drawn a lot of attention because of its high specific activity needed in radiolabelling studies. In this study, first, the radioactivity of 177Lu and other impurities that are produced during the production are calculated by MATLAB software. The experimental studies are followed by irradiating of enriched [176Y]Yb2O3 target in TRR in thermal neutron fluxes of 4
Keywords: lutethium-177; no carrier added; Tehran Research Reactor; MATLAB software; MCNPX2.6.
Vibration analysis of an irradiation device prototype for NTD-Si for the Es-Salam research reactor
by Mourad Dougdag, Mohammed Harek, Mhmed Salhi, Rachid Teberbek, Mouhammed Ouali
Abstract: The Neutron Transmutation Doping (NTD) technique of silicon is one of those sustainable applications in nuclear science. Providing such a product involves the construction of an irradiation device that meets quality requirements of doping, and addressing challenges related to the nuclear safety design. In the reactor core, the irradiation device will be subjected to different types of damage caused by radiation, thermal, pressure, dynamic loads, etc. This paper focuses on the assessment of harmful dynamic loads induced by the cooling air flow, the rotating driven system and by external hazards such as earthquakes to which the prototype device at reduced length, constructed in Birine Research Centre of Algeria, is subjected. The main goals are to perform vibration tests out-of-pile at normal operation conditions, to verify that the prototype respects compliance criteria, to select the best operating parameters and to develop and adjust a numerical model. The obtained model will be used to improve an extended full-scale one. The experimental tests and numerical models give globally satisfactory results. Thus, the design of the prototype irradiation device is considered reliable and safe, which allows the realisation of the expected real irradiation device.
Keywords: irradiation device prototype; Es-Salam research reactor; NTD-Si; vibration; FEM; EMA; model adjustment.
Investigations on the iso-geometric analysis method applied to neutron transport calculations in two- and three-Dimensional spatial domains
by Matthias Nezondet, Riho Horita, Willem Van Rooijen
Abstract: A simulation method for the transport of neutrons in spatial domains of arbitrary shape is presented, using $S_N$ neutron transport theory and the Iso-Geometric Analysis (IGA) method. IGA is a method to solve partial differential equations, such as the neutron transport equation, on spatial domains of arbitrary shape. A multi-group $S_N$ transport code capable of simulating two- and three-dimensional spatial domains is developed. Examples of the code validation are presented, as well as examples of calculations of neutron transport in spatial domains of arbitrary shape. The result is that the IGA method allows to reach good accuracy, but in comparison to legacy codes based on the $S_N$ method, the IGA method is expensive in terms of calculation time and memory footprint. The combination of $S_N$ transport and the IGA method opens a possibility for reference calculations of neutron transport in complex spatial domains.
Keywords: neutron transport theory; IGA method; SN method; nuclear reactor analysis; arbitrary geometry.