Results 11 to 20 of about 3,515 (225)
Brazilian Journal of Radiation Sciences, 2021 This work proposes to implement the cell calculation considering the fuel burning using the WIMSD-5B code. The cell calculation procedure allows determining the nuclear parameters present in the multi-group neutron diffusion equation and for this purpose
Desirée Yael de Sena Tavares +2 more
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VALIDATION OF SERPENT FOR FUSION NEUTRONICS ANALYSIS AT JET [PDF]
EPJ Web of Conferences, 2021 Fusion neutronics analysis before and after experiments at JET is traditionally performed using Monte Carlo particle transport code Monte Carlo N-Particle.
Žohar Andrej +5 more
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Evaluation of PWR pressure vessel fast neutron fluence benchmarks from NUREG/CR-6115 with ares transport code [PDF]
Nuclear Technology and Radiation Protection, 2017 An accurate evaluation of PWR pressure vessel fast neutron fluence is essential to ensure pressure vessel integrity over the design lifetime. The discrete ordinates method is one of the main methods to treat such problems.
Zhang Liang +3 more
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Nuclear Engineering and Technology, 2023 MgF2 as a moderator material has been extensively used in the beam shaping assembly (BSA) that plays an important role in the boron neutron capture therapy (BNCT).
Jiaqi Hu +8 more
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MC++: a parallel, portable, Monte Carlo neutron transport code in C++ [PDF]
Proceedings of 1997 SCS Simulation Multiconference, 2002 MC++ is a multi group Monte Carlo neutron transport code written in C++ and based on the Parallel Object Oriented Methods and Applications (POOMA) class library. MC++ runs in parallel on and is portable to a wide variety of platforms, including MPPs, SMPs, clustered SMPs, and clusters of Unix workstations.
Lee, S. R. +2 more
openaire +2 more sources
APPLICATION OF THE JSIR2S CODE PACKAGE FOR SHUTDOWN DOSE RATE CALCULATIONS ON JET [PDF]
EPJ Web of Conferences, 2021 In this paper we present a computational exercise for shut-down dose rate calculations for the JET tokamak using the in-house developed JSIR2S code package as part of its validation.
Ambrožič Klemen +3 more
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Advances for the time-dependent Monte Carlo neutron transport analysis in McCARD
Nuclear Engineering and Technology, 2023 For an accurate and efficient time-dependent Monte Carlo (TDMC) neutron transport analysis, several advanced methods are newly developed and implemented in the Seoul National University Monte Carlo code, McCARD.
Sang Hoon Jang, Hyung Jin Shim
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MULTI-PHYSICS TRANSIENT SIMULATIONS WITH TRIPOLI-4® [PDF]
EPJ Web of Conferences, 2021 In this work, we present the first dynamic calculations performed with the Monte Carlo neutron transport code TRIPOLI-4R with thermal-hydraulics feedback.
Faucher Margaux +2 more
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Parallel performance results for the OpenMOC neutron transport code on multicore platforms [PDF]
The International Journal of High Performance Computing Applications, 2016 The shift toward multicore architectures has ushered in a new era of shared memory parallelism for scientific applications. This transition has introduced challenges for the nuclear engineering community, as it seeks to design high-fidelity full-core reactor physics simulation tools.
Siegel, A. +4 more
openaire +3 more sources
Energies, 2023 Multigroup constants are the foundation of neutron and photon transport problems, and the accuracy of multigroup cross-sections has a significant impact on shielding calculation.
Xu Zhao +5 more
doaj +1 more source