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Showing 1 to 20 of 20 for “"neutron transport equation"”.
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A low order acceleration scheme for solving the neutron transport equation
… used technique for solving partial differential equations, and has been applied to the neutron transport problems for many years. The MOC method requires many transport iterations to solve large heterogeneous LWR reactor problems with high dominance ratio, and effective acceleration schemes are …
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Functional analytic treatment of linear transport equations in kinetic theory and neutron transport theory
The temperature-density equation of Kinetic Theory and the conservative neutron transport equation are studied. In both cases a modified version of the Larsen-Habetler resolvent integration technique is applied to obtain full-range and half-range expansions. For the neutron transport equation the …
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The Green's dyadic method for the analytic solution of multigroup neutron transport boundary value problems
… for the exact solution of the multigroup neutron transport equation. The method is based on the treatment of the steady-state, plane geometry multigroup equations for general anisotropic scattering in a homogenous medium. The problem of a system of N coupled integrodifferential equations …
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Transient method of characteristics via the Adiabatic, Theta, and Multigrid Amplitude Function methods
… (MOC) formulation of the time-dependent neutron transport equation. The transient transport versions of the 2D LRA and C5G7 benchmarks were used to assess the performance and accuracy of these methods. We began by deriving the CMFD-accelerated MOC algorithm in 2D steady state form and …
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The Asymptotic Development of The Equations of Nuclear Reactor Theory
The solutions to various forms of the diffusion equation are shown to correspond to the leading term of an asymptotic expansion of the solution to the neutron transport equation for large near-homogeneous systems which are close to critical. The product of the fundamental unit-cell transport theory …
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Transport-theory-equivalent diffusion coefficients for node-homogenized neutron diffusion problems in CANDU lattices
Calculation of the neutron flux in a nuclear reactor core is ideally performed by solving the neutron transport equation for a detailed-geometry model using several tens of energy groups. However, performing such detailed calculations for an entire core is prohibitively expensive from a …
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A novel equivalence method for high fidelity hybrid stochastic-deterministic neutron transport simulations
… Downstream highly parallelized 3D deterministic transport can then use those cross sections to compute accurate solutions at the full core scale. This hybrid computation scheme makes the most of both worlds to achieve fast and accurate reactor physics simulations. Among the few remaining …
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Leakage-corrected discontinuity factors for a second-generation Th-Pu pressure-tube SCWR.
The neutron flux throughout a reactor core should be calculated, ideally, by solving the neutron transport equation for a highly detailed geometric model of the core. Since this is computationally impractical, approximate node-homogenized models have historically been used whereby neutronic …
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FASPEC, a program to determine group constants for up to 47 groups in a fast neutron spectrum
… by a solution of the Infinite medium neutron transport equation. Flux contributions from inelastic scatter are included while those from neutron up-scatter are not. The energy spectrum considered is from 10 MeV to 0.625 eV. Required input is the atomic number density of each isotope, …
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Fission Multiplicity Distribution Sampling in MCNP6 Criticality Calculations
In the solution of the neutron transport equation, the k-effective eigenvalue is related to the average number of neutrons emitted in fission of the system. It can be shown that if the average number of neutrons emitted in fission and the average neutron energy spectrum is conserved, the …
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Development of MURR flux trap model for simulation and prediction of sample loading reactivity worth and isotope production
… the reactor core and flux trap area, solves the neutron transport equation and calculates the loading worth based on the Monte Carol method, proceeds with burnup and decay calculation, and predicts the requested isotope production. MCNP part of the MFTM model carries out neutron transport …
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Modeling Feedback Effects of Transient Nuclear Systems Using Monte Carlo
Monte Carlo neutron transport is the gold standard for accurate neutronics simulation of nuclear reactors in steady-state because each term of the neutron transport equation can be directly tallied using continuous-energy cross sections rather than needing to make approximations in energy, angle, …
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Novel Resonance Self-Shielding Methods for Nuclear Reactor Analysis
In the simulation of the behavior of neutrons in a nuclear reactor, there has long been a dichotomy in solution techniques. One can use Monte Carlo methods, known to be very accurate and problem agnostic but also very costly, or deterministic methods, known to be more computationally efficient but …
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Methods for including multiphysics feedback in Monte Carlo reactor physics calculations
… Carlo methods are desirable for solving the neutron transport equation required in reactor analysis because of the inherent accuracy of the method, but the Computational Solid Geometry (CSG) representation of the physical geometry makes it difficult to accurately and efficiently perform …
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Variable Fidelity Monte Carlo Neutron Transport: On Combining Multi-Group and Continuous Energy Representation in Monte Carlo Calculations
… investigates the feasibility of Monte Carlo (MC) neutronic calculations that combine multi-group (MG) and continuous energy (CE) representation of nuclear interaction data in different spatial regions. In general, MG MC calculations run 3-4 times faster than CE MC. Thus, it might be possible to …
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A reactor physics framework to detect anomalies in HTGR core
This dissertation extends the application of neutron noise analysis as a tool for detecting anomalies in High Temperature Gas-cooled Reactors (HTGRs). Neutron noise is defined as the fluctuations in the neutron detector signal caused by perturbations in a nuclear reactor. Neutron noise analysis is …
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A lattice boltzmann/s n framework for coupled heat transfer and neutron transport problems
… tight coupling — to solve coupled mathematical equations describing a multi-physics system. The major advantage of loose coupling is to allow separately validated codes to perform specific calculations without modifying them. This may be the only approach available when users do not have access …
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First-Flight Effects on Reactor Power Shapes and Neutron Leakage
Made available in DSpace on 2017-07-06T18:55:52Z (GMT). No. of bitstreams: 3 Polich_Timothy_J_MS.pdf: 24183567 bytes, checksum: 27f4e23c0cc79dba37c0833bbd66c606 (MD5) Polich_Timothy_J_MS_ABS.pdf: 453613 bytes, checksum: 2d8988ab0c50536bb28a23682aac29e3 (MD5) license.txt: 4813 bytes, checksum: …
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Methods for safety and stability analysis of nuclear systems
L'abstract è presente nell'allegato / the abstract is in the attachment