Zobrazeno 1 - 10
of 362
pro vyhledávání: '"openmc"'
Autor:
Ahmad Muzaki Mabruri, Ratna Dewi Syarifah, Indarta Kuncoro Aji, Artoto Arkundato, Nuri Trianti
Publikováno v:
JIF (Jurnal Ilmu Fisika), Vol 16, Iss 2 (2024)
Validation of Low-Cycle and Low-Particle OpenMC Simulation Codes for Neutronics Calculations has been conducted. This study validates OpenMC, an evolving open-source neutron analysis code. Validation of Low-Cycle and Low-Particle Codes is crucial as
Externí odkaz:
https://doaj.org/article/8fcd98ba9b5f4bc5b480af771be8b085
Publikováno v:
Frontiers in Nuclear Engineering, Vol 3 (2024)
In this study, we present a detailed comparison of two independently developed models of the Molten Salt Reactor Experiment (MSRE) for Monte Carlo particle transport simulations: the constructive solid geometry (CSG) model that was developed in suppo
Externí odkaz:
https://doaj.org/article/b4b64732bdb24aa98e67dc822bac31f4
Publikováno v:
Nuclear Energy and Technology, Vol 9, Iss 4, Pp 215-225 (2023)
The goal of this study is to perform neutronic calculations of the VVER-1000 MOX core computational benchmarks with an OpenMC code along with ENDF/B-VII.1 nuclear data library. The results of neutronic analysis using the OpenMC Monte Carlo code for t
Externí odkaz:
https://doaj.org/article/9c2156a3521c44d88d7feb69198ce9da
Autor:
J. Romero-Barrientos, F. Molina, J.I. Márquez Damián, M. Zambra, P. Aguilera, F. López-Usquiano, S. Parra
Publikováno v:
Nuclear Engineering and Technology, Vol 55, Iss 5, Pp 1593-1603 (2023)
In deterministic and Monte Carlo transport codes, β-delayed emission is included using a group structure where all of the precursors are grouped together in 6 groups or families, but given the increase in computational power, nowadays there is no re
Externí odkaz:
https://doaj.org/article/e1369940786d4e29a66ca44c1069c640
Autor:
Hamza A Tanash, Denis A Solovyov, Vyacheslav G Zimin, Alexey L Lobarev, Denis A Plotnikov, Nikolay V Schukin
Publikováno v:
Глобальная ядерная безопасность, Vol 0, Iss 1, Pp 79-91 (2023)
OpenMC is a state-of-the-art Monte Carlo neutron transport simulation code that uses the Python programming language as an API. OpenMC supports eight burnout simulation algorithms. This study presents the results of choosing an integration method for
Externí odkaz:
https://doaj.org/article/efdb8dfdc929436a8e8144834dc36fc8
Publikováno v:
Energies, Vol 17, Iss 9, p 2177 (2024)
The computational demand of neutron Monte Carlo transport simulations can increase rapidly with the spatial and energy resolution of tallied physical quantities. Convolutional neural networks have been used to increase the resolution of Monte Carlo s
Externí odkaz:
https://doaj.org/article/e3dfedca39274481ae9b3a6ecd41a8df
Inter-Code Comparison of Computational VERA Depletion Benchmark Using OpenMC, OpenMC-ONIX and DRAGON
Publikováno v:
Atom Indonesia, Vol 48, Iss 3, Pp 193-203 (2022)
This research focuses on the comparative analysis of the PWR fuel assembly based on VERA depletion benchmark problems using community-developed open source Monte Carlo code OpenMC, python based burnup code system ONIX (a coupling interface for Monte
Externí odkaz:
https://doaj.org/article/7d3764e9bb444104a82ebbf318f38f17
Autor:
Khurram Mehboob, Yahya A. Al-Zahrani
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 12, Pp 4571-4584 (2022)
The analysis of rapid flow transients in Reactor Coolant Pumps (RCP) is essential for a reactor safety study. An accurate and precise analysis of the RCP coastdown is necessary for the reactor design. The coastdown of RCP affects the coolant temperat
Externí odkaz:
https://doaj.org/article/0601ca4b6d834580aa6a714dc2ed01d9
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 10, Pp 3803-3810 (2022)
OpenMC is a community-driven open-source Monte Carlo neutron and photon transport simulation code. The Weight Window Mesh (WWM) function and an automatic Global Variance Reduction (GVR) method was recently developed and implemented in a developmental
Externí odkaz:
https://doaj.org/article/4c4b6a64b436469fbc4448a03ca3c343
Publikováno v:
Nuclear Engineering and Technology, Vol 54, Iss 10, Pp 3888-3896 (2022)
Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results
Externí odkaz:
https://doaj.org/article/6fb42e087d9d407cae43886feeef9bf3