Background: Recently, different investigators have studied the possibility of radiation therapy in restenosis prevention and have shown promising results. In this study a unique radioactive source for intra vascular brachytherapy (IVBT) was investigated. The two-dimensional dose distribution in water for a 32P IVBT stent has been calculated. The pure beta emitter source 32P has been co...

MCNP is the Monte Carlo N-Particle radiation transport code whose history dates back more than half a century to the early days of computing. From a simple beginning, its uses have grown to include fields such as criticality safety, radiation shielding, oil well logging, and medical imaging and diagnostics and an international user community of over 3000 users. This large user community could o...

For the first time, measurements of the time-dependent cross-correlation distributions of plutonium oxide have been made separately for neutrons and gamma rays. Six EJ-309 liquid scintillation detectors with a digital, offline pulse shape discrimination and pulse timing method were used to measure five different samples of varying mass and burnup. The number of (neutron, neutron) correlations w...

The axial distribution of energy deposited in bremsstrahlung conversion targets is modelled using the Monte Carlo N -Particle (MCNP) code. Systems comprising a disc-shaped target, segmented axially, and a point source of paraxial, monoenergetic electrons (2, 7, 10, or 14 MeV) are considered. Thick tantalum and aluminium targets and also tantalum targets of optimum thickness are modelled. Energy...

We present new results on neutron and gamma-ray pulse-height distributions (PHDs) measured with liquid scintillators from five plutonium-oxide samples of varying mass and burnup and a Cf isotopic source. We show that the analysis of the pulse-height distributions can be used to easily distinguish the fissile material (plutonium oxide) from the Cf source. Moreover, the slope of the measured puls...

Detailed 3-D neutronics calculations have been performed for the US DCLL TBM. The neutronics calculations were performed directly in the CAD model using the DAG-MCNP code that allows preserving the geometrical details. Detailed high-resolution, high-fidelity profiles of the nuclear parameters were generated using fine mesh tallies. These included tritium production, nuclear heating, and radiati...

Background: Monte Carlo determination of TG-43 brachytherapy dosimetry parameters and dose distribution calculation for 131Cs source model CS-1 are presented in this study. Materials and Methods: The dose distribution was calculated around the 131Cs Model CS-1 located in the center of 30 cm ×30 cm ×30 cm water, and soft tissue phantoms cube using MCNP code by Monte Carlo method. The percentage ...

The MCS code is a computer developed by the Ulsan National Institute of Science and Technology (UNIST) for simulation calculation nuclear reactor systems based on Monte Carlo method. currently used to solve two main types physics problems, namely, criticality problems radiation shielding problems. In this paper, capability validated simulating some selected SINBAD (Shielding Integral Benchmark ...