Results 61 to 70 of about 538 (156)

Validasi Kode OpenMC pada Reaktor Gas Berpendingin Helium Berbahan Bakar UC-PuC

, 2023
Validasi perhitungan kekritisan pada Gas Cooled Fast Reactor (GFR) menggunakan kode OpenMC dan SRAC telah dilakukan. OpenMC merupakan kode analisis neutronik yang bersifat open source dan probabilistik yang sedang dikembangkan oleh MIT hingga sekarang ...
Maulina, Wenny, Hidayat, Umar Sahiful, Rohman, Lutfi, Karomah, Iklimatul, Syarifah, Ratna Dewi, Purwandari, Endhah, Trianti, Nuri, Arkundato, Artoto   +7 more
core  

OpenMC In Situ Source Convergence Detection

, 2016
We designed and implemented an in situ version of particle source convergence for the OpenMC particle transport simulator. OpenMC is a Monte Carlo based-particle simulator for neutron criticality calculations. For the transport simulation to be accurate, source particles must converge on a spatial distribution.
Garrett Aldrich, Soumya Dutta, Jonathan Woodring   +2 more
openaire   +2 more sources

OpenMC data for simulating ARC reactor blanket

, 2020
The following datasets are the main OpenMC output files converted from dataframe to xlsx files. They are divided in the enrichment folder, where each salt has been parametrized over Li-6 enrichment fraction, and the mesh folder, where mesh results have ...
Segantin, S (via Mendeley Data)
core   +2 more sources

Coupled neutronic-thermal-mechanical simulation of the KRUSTY heat pipe microreactor

Frontiers in Nuclear Engineering
Multiphysics analysis has become a common technique for nuclear reactor design validation, with neutronic-thermal analysis being the typical choice for understanding reactor dynamics.
William Reed Kendrick, Benoit Forget
doaj   +1 more source

Benchmarking of the SPERT-III E-core experiment with the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC [PDF]

EPJ Web of Conferences
This paper presents a code-to-code verification of the SPERT-III reactor in its E-core configuration using the Monte Carlo codes TRIPOLI-4®, TRIPOLI-5® and OpenMC.
Dervaux Antoine, Kinast Shai, Lemaire Matthieu, Di Blasi Vincenzo, Larmier Coline, Mancusi Davide, Zoia Andrea   +6 more
doaj   +1 more source

Perhitungan Shutdown Margin Teras NuScale Menggunakan OpenMC

, 2023
One of the important reactor safety parameters to study is the issue of shutdown margin (SDM). This study aims to obtain an effective safety design of the NuScale reactor in reviewing the SDM value parameter.
Razaqiyanto, Hadi, Hadi Razaqiyanto, Shafii, Mohammad Ali, Raflis, Helen, Mohammad Ali Shafii, Helen Raflis   +5 more
core   +2 more sources

Using OpenMC in simulations of a low dimensional cold neutron moderator for the ICONE project [PDF]

EPJ Web of Conferences
In this contribution we showcase the use of OpenMC for the validation and optimization of a low dimensional cold neutron moderator for the ICONE project. An overview of our simulation strategy is given and initial results for an optimized layout of a low-
Wagner Richard, Ott Frédéric, Darpentigny Jacques   +2 more
doaj   +1 more source

ANALISIS REAKTIVITAS BATANG KENDALI SMALL MODULAR PWR MENGGUNAKAN OPENMC [PDF]

, 2023
Salah satu parameter keselamatan reaktor yang penting untuk dipelajari adalah masalah reaktivitas lebih (core excess), dan shutdown margin (SDM). Penelitian ini bertujuan untuk melihat nilai reaktivitas dan SDM dari pengaruh penarikan batang kendali ...
Hadi, Razaqiyanto
core  

Methodology and preliminary verification of generating heterogeneous multigroup microscopic cross-section libraries for neutron transport codes based on OpenMC

Nuclear Engineering and Technology
Advancements in reactor technology, particularly Generation IV and modular reactors, have introduced new challenges on the neutronics analysis due to their complex geometries and spectra.
Bowen Cui, Guohua Chen, Xiaofeng Jiang
doaj   +1 more source

Extension of the OpenMC depletion module for transport-independent depletion

Proceedings of the Python in Science Conference
This paper describes new functionality in OpenMC's depletion module for depleting materials independent of a neutron transport simulation. The paper valdiates the capability against transport-coupled depletion on a simple model.
Oleksandr Yardas, Paul Romano, Madicken Muk, Kathryn Huff   +3 more
openaire   +1 more source