Results 1 to 10 of about 3,444 (154)
AIP Advances, 2022 Accelerator-based neutron sources have been developed and installed in recent decades for boron neutron capture therapy (BNCT) in several clinical facilities.
Yusuke Matsuya +9 more
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Analysis of Kobayashi benchmark with indigenous Monte Carlo neutron transport code PATMOC
Physics Open, 2023 The solution of the neutron transport equation is the basic input for the reactor physics design of a nuclear reactor system. Because of the complexities of geometry and cross-section data, the neutron transport equation is generally solved using ...
Amod Kishore Mallick, Umasankari Kannan
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TOUCANS: A versatile Monte Carlo neutron transport code based on GEANT4
Nuclear Instruments and Methods in Physics Research, Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 2023 17 pages, 9 ...
Eric Dumonteil +2 more
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Hybrid parallel strategies for the neutron transport code PANDAS-MOC
Frontiers in Nuclear Engineering, 2022 PANDAS-MOC (Purdue Advanced Neutronics Design and Analysis System with Methods of Characteristics) is being developed to find high fidelity 3D solutions for steady state and transient neutron transport analysis. However, solving such transport problems in a large reactor core could be extremely computationally intensive and memory demanding.
Yunlin Xu, Shunjiang Tao
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Status of the SCONE Monte Carlo neutron transport code
Annals of Nuclear EnergySCONE is a Monte Carlo neutron transport code written at the University of Cambridge. It is the first high-performance Monte Carlo code written in the United Kingdom since MONK and is Britain’s only open-source neutron transport code. As well as having performance and accuracy representative of modern Monte Carlo codes, SCONE is distinguished by its ...
Valeria Raffuzzi +2 more
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Deterministic neutron transport and burnup code coupling assessment
Annals of Nuclear EnergyThis work has been partially supported by Grant PGC2018-096437-B-I00-AR funded by MCIN/AEI/10.13039/501100011033 and by "ERDF A way of making Europe", by the "European Union" and Grant PRE2019-089431 funded by MCIN/AEI/10.13039/501100011033 and by "ESF Investing in your future".
T Barrachina, Rafael Miro
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Parallel schedules for MOC sweep in the neutron transport code PANDAS-MOC
Frontiers in Nuclear Engineering, 2022 Method of characteristics (MOC) is a commonly applied technique for solving the Boltzmann form of the neutron transport equation. In the PANDAS-MOC neutron transport code, MOC is used to determine the 2D radial solution. However, in the whole-code OpenMP threading hybrid model (WCP) of PANDAS-MOC, it is found that when using the classic parallelism ...
Yunlin Xu, Shunjiang Tao
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Trudy Odesskogo Politehničeskogo Universiteta, 2017 The constant preparation issues are important for performing the few-group analysis for different states of the reactor core. Besides, the constant preparation method influences the further calculations accuracy and quality.
Віталій Володимирович Гальченко +2 more
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Nuclear Engineering and Technology, 2022 In this paper, a new multi-group neutron-gamma transport calculation code system STRAUM-MATXST for complicated geometrical problems is introduced and its development status including numerical tests is presented.
MyeongHyeon Woo, Ser Gi Hong
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ANALYSIS AND COMPARISON OF APOLLO3® AND TRIPOLI-4® NEUTRON NOISE SOLVERS [PDF]
EPJ Web of Conferences, 2021 Neutron noise analysis addresses the description of small time-dependent flux fluctuations in reactor cores, induced by small global or local perturbations of the macroscopic cross sections due to density fluctuations of the coolant, to vibrations of ...
Rouchon Amélie +2 more
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