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Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results), chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

The objective of this paper is to introduce a new direct method for neutronic calculations. This method, called direct discrete method, is simpler than the application of the neutron transport equation and more compatible with the physical meanings of the problem. The method, based on the physics of the problem, initially runs through meshing of the desired geometry. Next, the balance equation for each mesh interval is written. Considering the connection between the mesh intervals, the final discrete equation series are directly obtained without the need to pass through the set up of the neutron transport differential equation first. In this paper, one and multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without the associated clad and the coolant regions each with two different external boundary conditions. The validity of the results from this new method is tested against the results obtained by the MCNP-4B and the ANISN codes.

In this study, after discretization of the neutron diffusion equation and adjoint with high-order nodal expansion method in two dimensions and two energy groups, calculations with Momentum and Galerkin weighting functions for rectangular geometry (BIBLIS-2D) and hexagonal geometry (IAEA-2D) reactors are performed. The mean of relative power error for Momentum and Galerkin weighting functions was calculated in BIBLIS-2D reactor 0.42% and 0.62%, respectively, and for IAEA-2D reactor 4.96% and 3.52%, respectively. Regarding the results, it was concluded that in order to increase accuracy with the acceptable time of computing (4 Seconds for rectangular geometry and 28 seconds for hexagonal geometry with Intelr CoreTM i7-4510U Processor), the Momentum weighting function for rectangular geometry and the Galerkin weighting function for hexagonal geometry can be used to discretize equations without reducing the node size. Therefore, to increase the accuracy while maintaining the speed of calculations, without reducing the size of nodes, the appropriate weighting function can be used in discretization, which can be very useful in performing calculations of different transients. [ABSTRACT FROM AUTHOR]

In this study, optimized preparation, quality control, cell assessments and biostribution of 188Re-HYNIC-PSMA in normal rats and tumor bearing mice are studied. Optimized conditions for radiolabeling were evaluated and radiochemical purity (> 99%) was investigated using ITLC and HPLC methods. 188Re-HYNIC-PSMA was stable both in PBS (4 °C) and in human serum (37 °C) even after 48 h. The results showed the complex was cleared from the blood very fast via urinary track. This new radiolabeled compound has a high potential to be considered as an agent for the treatment of patients with PSMA expressing tumors however more biological studies are still needed. [ABSTRACT FROM AUTHOR]