Radiation damage and yield strength calculations in the pressure vessel of WWER-1000 reactor and sensitivity analysis of neutron spectrum on the shield of vessel [PDF]
مجله علوم و فنون هستهای, 2021 Radiation damage calculation is very important for estimating the lifetime of nuclear power plant instruments. The purpose of the present study is to evaluate the sensitivity of the damage values to the neutron group energy, where the neutron flux rate ...
E. Moslemi-Mehni +3 more
doaj +1 more source
Monte Carlo domain decomposition for robust nuclear reactor analysis [PDF]
, 2014 Monte Carlo (MC) neutral particle transport codes are considered the gold-standard for nuclear simulations, but they cannot be robustly applied to high-fidelity nuclear reactor analysis without accommodating several terabytes of materials and tally data.
Forget, Benoit Robert Yves +3 more
core +1 more source
Progress toward Monte Carlo–thermal hydraulic coupling using low-order nonlinear diffusion acceleration methods [PDF]
, 2017 A new approach for coupled Monte Carlo (MC) and thermal hydraulics (TH) simulations is proposed using low-order nonlinear diffusion acceleration methods.
Forget, Benoit Robert Yves +2 more
core +1 more source
Optimizations of the energy grid search algorithm in continuous-energy Monte Carlo particle transport codes [PDF]
, 2017 In this work we propose, implement, and test various optimizations of the typical energy grid-cross section pair lookup algorithm in Monte Carlo particle transport codes.
Forget, Benoit Robert Yves +3 more
core +1 more source
OpenMC: Towards Simplifying Programming for TianHe Supercomputers
Journal of Computer Science and Technology, 2014 Modern petascale and future exascale systems are massively heterogeneous architectures. Developing productive intra-node programming models is crucial toward addressing their programming challenge. We introduce a directive-based intra-node programming model, OpenMC, and show that this new model can achieve ease of programming, high performance, and the
Liao, XK +6 more
openaire +2 more sources
Upsampling Monte Carlo neutron transport simulation tallies using a convolutional neural network
Energy and AI, 2023 The physical quantities calculated by nuclear reactor Monte Carlo simulations are typically recorded on a grid of two or three spatial dimensions and one dimension of neutron energy. Because of this, increasing the resolution of the calculated quantities
Andrew Osborne +2 more
doaj +1 more source
On the use of tally servers in Monte Carlo simulations of light-water reactors [PDF]
, 2014 An algorithm for decomposing tally data in Monte Carlo simulations using servers has recently been proposed and analyzed. In the present work, we make a number of refinements to a theoretical performance model of the tally server algorithm to better ...
Brown +6 more
core +1 more source
ANALYSIS OF CORE CONFIGURATION FOR CONCEPTUAL GAS COOLED FAST REACTOR (GFR) USING OPENMC [PDF]
, 2023 This study focused on a conceptual core configuration of Gas Cooled Fast Reactor (GFR), as part of a generation IV reactor. Uranium-plutonium carbide (UC-PuC) was used as reactor fuel and a Monte Carlo simulation method using OpenMC has been carried out.
Karomah, Iklimatul +3 more
core +2 more sources
Data decomposition of Monte Carlo particle transport simulations via tally servers [PDF]
, 2013 An algorithm for decomposing large tally data in Monte Carlo particle transport simulations is developed, analyzed, and implemented in a continuous-energy Monte Carlo code, OpenMC.
Forget, Benoit Robert Yves +3 more
core +1 more source
Progress and Status of the Openmc Monte Carlo Code [PDF]
SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo, 2017 United States. Department of Energy. Naval Reactors Division (Rickover Fellowship Program in Nuclear Engineering)
Siegel, Andrew R. +5 more
openaire +5 more sources