Efficient burnup of minor actinides is one of the most promising alternatives for minimizing waste in advanced nuclear fuel cycles. This work examines the concept of employing a Z-pinch driven fusion source in a sub-critical transmutation reactor designed to burn up actinides and generate constant power. Its fuel cycle is designed to allow on-line fission product removal and fuel replenishment....

The problem of monoenergetic neutral particle transport in a duct, where particles travel inside the duct walls, is treated using an approximate one-dimensional model. The onedimensional model uses three-basis functions, as part of a previously derived weighted-residual procedure, to account for the geometry of particle transport in a duct system (where particle migration into the walls is not ...

The MCNP Monte Carlo radiation transport code was used for calculation of scatter distribution and scatter-toprimary ratio (SPR) for different tube voltages, phantom thicknesses and field size in diagnostic radiology. Subsequently, the effect of grid’s design parameters such as strip density, grid ratio, interspace material and lead-tointerspace ratio on scatter rejection was investigated by ca...

This paper presents an accurate and efficient approach to optimize radiation transport simulations in a stochastic medium of high heterogeneity, like the Very High Temperature Gas-cooled Reactor (VHTR) configurations packed with TRISO fuel particles. Based on a fast nearest neighbor search algorithm, a modified fast Random Sequential Addition (RSA) method is first developed to speed up the gene...

Theoretical and experimental dosimetric studies have been supplied useful information on the dependence of the brachytherapy source geometry and materials (1-5). Usually, Monte Carlo method is used to define dose distribution function, the radial dose variation, and the dose calculation close to the source in brachytherapy. Chen et al. (6). calculated the distribution of absorbed dose around co...

The general purpose Monte Carlo N-particle radiation transport computer code (MCNP4C) was used for the simulation of x-ray spectra in diagnostic radiology and mammography. The electrons were transported until they slow down and stop in the target. Both bremsstrahlung and characteristic x-ray production were considered in this work. We focus on the simulation of various target/filter combination...

To account for particle interactions in the extracellular (physical) environment, information from the cell-level Monte Carlo damage simulation (MCDS) for DNA double strand break (DSB) induction has been integrated into the general purpose Monte Carlo N-particle (MCNP) radiation transport code system. The effort to integrate these models is motivated by the need for a computationally efficient ...

In this paper, an absolute measurements technique for the sub-criticality determination is presented. The development of ADS, requires of methods to monitor and control the sub-criticality of this kind of systems, without interfering it’s normal operation mode. This method is based on the Stochastic Neutron and Photon Transport Theory developed by Muñoz-Cobo et al. [1], and which can be impleme...

The Neutron induced activity and decay heat are very important for the nuclear design of fusion reactor. In this paper, activation analysis of DEMO reactor with helium-cooled solid blanket (HCSB-DEMO) was performed. Ceramics Li4SiO4 and Beryllium options were considered as tritium breeding material and neutron multiplier, respectively. Chinese low-activation ferritic steel (CLF-1) developed in ...

Accuracy of treatment planning systems may significantly influence the efficacy of brachytherapy. The purpose of this work is a detailed, varied and independent evaluation of an in-house brachytherapy treatment planning software called STPS. Operational accuracy of STPS was investigated. Geometric tests were performed to validate entry and reconstruction of positional information from scanned o...