Results 51 to 60 of about 898 (171)

Fuel Performance Comparison of Uranium Nitride and Uranium Carbide in VVER-1200 using OpenMC [PDF]

MATEC Web of Conferences
Nuclear power is a reliable and large-scale source of GHG-free electricity. This study asses the viability of ATF fuel of uranium nitride (UN) and uranium carbide (UC) as fuel for the VVER-1200 reactor.
Jamil Meekal, Ali Majid
doaj   +1 more source

Analysis of correlations and their impact on convergence rates in Monte Carlo eigenvalue simulations [PDF]

, 2018
This paper provides an analysis of the generation-to-generation correlations as observed when solving full core eigenvalue problems on PWR systems. Many studies have in the past looked at the impact of these correlations on reported variance and this ...
Forget, Benoit Robert Yves, Miao, Jilang, Smith, Kord S.   +2 more
core   +1 more source

Extending OpenMC validation to spent fuel canisters: A criticality benchmark against MCNP

Nuclear Engineering and Technology
OpenMC is an open-source Monte Carlo code with increasing relevance in criticality safety and reactor physics applications. While its validation has covered a broad range of systems, its performance in spent nuclear fuel storage scenarios remains limited
J. Ruiz-Pineda, J. Romero-Barrientos, F. Molina, M. Zambra, F. López-Usquiano   +4 more
doaj   +1 more source

ANALISIS KRITIKALITAS PROSES HYDRIDING SERTA PENYIMPANAN PADUAN UMo DAN UZr SEBAGAI KANDIDAT BAHAN BAKAR REAKTOR RISET MENGGUNAKAN OPENMC

Urania, 2023
Penelitian bahan bakar reaktor riset saat ini mengarah pada peningkatan densitas uranium untuk menghasilkan fluks neutron yang lebih tinggi dalam rangka peningkatan produksi radioisotop serta  efisiensi dan akurasi pada berbagai pengujian material ...
Saga Octadamailah, Anissa Isnaini, Rohmad Sigit   +2 more
doaj   +1 more source

The OpenMOC method of characteristics neutral particle transport code [PDF]

, 2013
The method of characteristics (MOC) is a numerical integration technique for partial differential equations, and has seen widespread use for reactor physics lattice calculations.
Boyd III, William Robert Dawson, Forget, Benoit Robert Yves, Li, Lulu, Shaner, Samuel Christopher, Smith, Kord S.   +4 more
core   +1 more source

Multiphysics Analysis of Natural Circulation‐Driven Operation of Passive Molten Salt Fast Reactor and Effect of Guide Structure

International Journal of Energy Research, Volume 2025, Issue 1, 2025.
A new molten salt reactor (MSR) design has been developed aiming for long‐term operation and high safety. In order to enhance the integrity and economy of the system during the long‐term operation, pumps were removed from the primary system, and the fuel salt flow was developed by natural circulation.
Juhyeong Lee, Min Seop Song, Seungug Jae, Won Sik Yang, Sung Joong Kim, Mahmoud Ahmed   +5 more
wiley   +1 more source

A Cumulative migration method for computing rigorous transport cross sections and diffusion coefficients for LWR lattices with Monte Carlo [PDF]

, 2016
A new method for computing homogenized assembly neutron transport cross sections and diffusion coefficients that is both rigorous and computationally efficient is proposed in this paper.
Forget, Benoit Robert Yves, Liu, Zhaoyuan, Smith, Kord S.   +2 more
core  

Evaluation of Improved Contribution Function Method in the Compression of CINDER90 Depletion Library

Science and Technology of Nuclear Installations, Volume 2024, Issue 1, 2024.
The improved contribution function (ICF) method based on generalized perturbation theory (GPT) is applied to the compression of CINDER90 depletion library. A series of representative problems for PWR are defined with different fuel materials and operation conditions.
Yunfei Zhang, Qian Zhang, Qiang Zhao, Yang Zou, Jian Guo, Rui Yan, Guifeng Zhu, Arkady Serikov   +7 more
wiley   +1 more source

Petri Net Reachability Graphs: Decidability Status of FO Properties [PDF]

, 2011
We investigate the decidability and complexity status of model-checking problems on unlabelled reachability graphs of Petri nets by considering first-order, modal and pattern-based languages without labels on transitions or atomic propositions on ...
, Darondeau, Philippe, Meyer, Roland, Morvan, Christophe   +3 more
core   +4 more sources

The Study of Continuous Core Zoning to Extend the Graphite Component Irradiation Lifespan in Molten Salt Reactor

International Journal of Energy Research, Volume 2024, Issue 1, 2024.
Graphite is widely used in molten salt reactors (MSRs) because of its excellent properties. However, the irradiation lifespan of graphite in MSR is much shorter than the life of a conventional nuclear power plant, which lowers the load factor and increases the economic burden.
Shuyang Jia, Guifeng Zhu, Yang Zou, Jian Guo, Zhenghao Xu, Siqin Hu, Hongjie Xu, Sanjay Basumatary   +7 more
wiley   +1 more source