In this work we present a predictive analytical model that encompasses the performance and scaling characteristics of a non-deterministic particle transport application, MCNP. Previous studies on the scalability of parallel Monte Carlo eigenvalue calculations have been rather general in nature [1]. It can be used for the simulation of neutron, photon, electron, or coupled transport, and has fou...

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...

The Monte Carlo calculation of the effective (i.e. adjoint weighted) neutron generation time Keff, especially for continuous energy simulations, is not straightforward nor standard in Monte Carlo codes. The use of the non-adjoint weighted neutron generation time K (standard in most Monte Carlo codes) as an approximation of the effective one can lead to a serious bias. We show here that the diff...

Epithermal neutron resonance self-shielding factors in foils of materials used as activation detectors or as targets for radionuclide production have been calculated using the MCNP code. Two irradiation conditions have been considered: (a) foils immersed into an isotropic neutron flux and (b) foils submitted to a collimated neutron beam. The self-shielding factors, Gres(t), for gold, indium, ma...

Lung brachytherapy using high-dose rate (192)Ir technique is a well-established technique of radiation therapy. However, many commercial treatment planning systems do not have the ability to consider the inhomogeneity of lung in relation to normal tissue. Under such circumstances dose calculations for tissues and organs at risk close to the target are inaccurate. The purpose of the current stud...

In the present work, a graphical user interface is developed to automate in-core fuel management using MCODE, an MCNP-ORIGEN linkage code. Data abstraction is achieved by means of five object classes that define the run, fuel assembly locations, fuel assemblies, fuel paths, and materials. The GUI and an associated fuel management wrapper were developed in Python, with the PyQt extension being u...

This work presents experimental dosimetry results for a new 103Pd brachytherapy seed, in accordance with the AAPM TG-43U1 recommendation that all new low-energy interstitial brachytherapy seeds should undergo one Monte Carlo (MC) and at least one experimental dosimetry characterization. Measurements were performed using TLD-GR200A circular chip dosimeters using standard methods employing thermo...

The purpose of this work carried out by MTA KFKI Atomic Energy Research Institute in Hungary, is to develop and validate a method for predicting the ex-core detector responses in the Russian pressurized water type VVER reactors. The time and space dependent neutron flux in the core during these measurements have been calculated by the KIKO3D nodal diffusion code. For calculating the ionization ...

This paper presents the validation of the Monte Carlo code MCNP-PoliMi for the simulation of nuclear safeguards experiments with plutonium and uranium metal based on fast time-correlation measurements. A comparison is presented between experimental data acquired with the Nuclear Materials Identification System and the Monte Carlo simulations. The measurements and simulations were performed for ...

introduction: average glandular dose calculation in mammography with mo-rh target-filter and dose calculation for different situations is accurate and fast. material and methods: in this research, first of all, x-ray spectra of a mo target bombarded by a 28 kev electron beam with and without a rh filter were calculated using the mcnp code. then, we used the sobol-wu parameters to write a fortra...