Results 21 to 30 of about 12,472 (163)

Variational nodal methods for neutron transport: 40 years in review

Nuclear Engineering and Technology, 2022
The variational nodal method for solving the neutron transport equation has evolved over 40 years. Based on a functional form of the Boltzmann neutron transport equation, the method now comprises a complete set of variants that can be employed for ...
Tengfei Zhang, Zhipeng Li
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Multi-species Neutron Transport Equation [PDF]

Journal of Statistical Physics, 2019
The Neutron Transport Equation (NTE) describes the flux of neutrons through inhomogeneous fissile medium. Whilst well treated in the nuclear physics literature (cf. [9, 27]), the NTE has had a somewhat scattered treatment in mathematical literature with a variety of different approaches (cf. [8, 25]).
Alexander M. G. Cox, Simon C. Harris, Emma L. Horton, Andreas E. Kyprianou   +3 more
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Skin effect in neutron transport theory [PDF]

Physical Review E, 2021
We identify a neutron-flux "skin effect" in the context of neutron transport theory. The skin effect, which emerges as a boundary layer at material interfaces, plays a critical role in a correct description of transport phenomena. A correct accounting of the boundary-layer structure helps bypass computational difficulties reported in the literature ...
E. L. Gaggioli, D. M. Mitnik, O. P. Bruno   +2 more
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lp-CMFD acceleration schemes in multi-energy group 2D Monte Carlo transport

Frontiers in Energy Research, 2022
The linear prolongation flux update scheme is extended to both regular CMFD acceleration, as well as partial CMFD acceleration in 2D multi energy group Monte Carlo k-eigenvalue neutron transport problems.
Y. R. Than, S. Xiao
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3-D Neutron Transport Benchmarks [PDF]

Journal of Nuclear Science and Technology, 1991
Three-dimensional (3-D) neutron transport benchmark problems proposed from Osaka University to NEACRP in 1988 have been calculated by many participants and the results have been summarized. The results of k eff, control rod worth, and region-averaged group fluxes for proposed four core models calculated by various 3-D transport codes have been compared.
Toshikazu TAKEDA, Hideaki IKEDA
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Neutron Transport Simulations of RBMK Fuel Assembly Using Multigroup and Continuous Energy Data Libraries within the SCALE Code

Science and Technology of Nuclear Installations, 2021
The neutron transport simulations of RBMK-1500 fuel assembly were performed using both multigroup and continuous energy data libraries available within the SCALE code system in order to validate its suitability for the estimation of RBMK neutronic ...
Andrius Slavickas, Raimondas Pabarčius, Aurimas Tonkūnas, Sigitas Rimkevičius   +3 more
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Results of the reactor dosimetry experiments performed for verification of the neutron transport calculations at the Hungarian Paks Nuclear Power Plant [PDF]

EPJ Web of Conferences
The operational lifespan of all four units at the Hungarian Paks Nuclear Power Plant (NPP) has been extended by 20 years beyond the original 30-year term. This extension has been implemented gradually, in one- to two-year increments, since 2013.
Zsolnay Eva M., Czifrus Szabolcs, Kis Dániel, Horváth András, Gerényi Anita   +4 more
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Development of neutron diffusion and transport algorithms based on finite volume method

He jishu
BackgroundWith the development of nuclear-thermal coupling technology, it is essential to consider the strong coupling effects between multiple physics fields and achieve high precision and large-scale parallel computing.
LI Wei, LIU Xiaojing, CHAI Xiang, ZHAO Pengcheng   +3 more
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Invariant Imbedding and Neutron Transport Theory III-Neutron-Neutron Collision Processes [PDF]

Indiana University Mathematics Journal, 1959
Abstract : The effects on criticality of neutron-neutron collisions involving annihilation are investigated for one-dimensional, single and multi-group cases. The analytic treatment shows that regardless of the magnitude of the cross section for collision between moving neutrons, there is no critical length (mass).
Bellman, Richard, Kalaba, Robert, Wing, G. Milton   +2 more
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Accelerated solution methods for self-adjoint angular flux neutron transport equations based on scattering matrix decoupling

Nuclear Engineering and Technology
Advanced microreactors exhibit compact geometries, elevated neutron leakage, and pronounced anisotropic scattering. These characteristics necessitate high-fidelity neutron transport methods to accurately resolve core flux distributions.
Duoyu Jiang, Peng Xu, Xinbiao Jiang, Tianliang Hu, Lipeng Wang, Lu Cao, Huaqi Li, Da Li, Lixin Chen   +8 more
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