An analysis of condensation errors in multi-group cross section generation for fine-mesh neutron transport calculations

Boyd, William; Gibson, Nathan; Forget, Benoit; Smith, Kord

doi:10.1016/j.anucene.2017.09.052

Cited by 22 publications

(8 citation statements)

References 6 publications

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“…Many detailed treatments of this formulation are found in the literature. 4,9,20,43,46 This section gives an overview of the characteristic form of the steady-state Boltzmann neutron transport equation as used by MOCkingbird. Sections IV.A through IV.F start from the multigroup, isotropic scattering approximation and review the reduction to a problem that is discrete in both angle and space.…”

Section: Moc In Mockingbirdmentioning

confidence: 99%

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Method of Characteristics for 3D, Full-Core Neutron Transport on Unstructured Mesh

et al. 2021

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A myriad of nuclear reactor designs are currently being considered for next-generation power production. These designs utilize unique geometries and materials and can rely on multiphysics effects for safety and operation. This work develops a neutron transport tool, MOCkingbird, capable of three-dimensional (3D), full-core reactor simulation for previously intractable geometries. The solver is based on the method of characteristics, utilizing unstructured mesh for the geometric description. MOCkingbird is built using the MOOSE multiphysics framework, allowing for straightforward linking to other physics in the future. A description of the algorithms and implementation is given, and solutions are computed for two-dimensional/3D C5G7 and the Massachusetts Institute of Technology BEAVRS benchmark. The final result shows the application of MOCkingbird to a 3D, full-core simulation utilizing 1.4 billion elements and solved using 12 000 processors.

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Section: Moc In Mockingbirdmentioning

confidence: 99%

“…Graphics processing units have also been used to parallelize MOC (Ref. 46), which hold their own parallelization challenges.…”

Section: Ive Parallelism and Domain Decompositionmentioning

confidence: 99%

Method of Characteristics for 3D, Full-Core Neutron Transport on Unstructured Mesh

et al. 2021

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“…Following the multi-group assumption, another assumption made is that a large and complex problem can be broken up into small constant cross section regions i [9], and that these regions have group dependent isotropic sources (fission + scattering), Q i,g . Anisotropic sources are also possible with MOC based methods [10], but are not used here for simplicity.…”

Section: Derivationmentioning

confidence: 99%

A Comparison of Stochastic Mesh Cell Volume Computation Strategies for the Random Ray Method of Neutral Particle Transport

et al. 2021

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The random ray method is a recently developed neutron transport method that can be used to perform efficient full-core, general-purpose, high-fidelity 3D simulations of nuclear reactors. While Tramm et al. have so far documented the new random ray algorithm in several publications, one critical detail has not yet been published: how to best determine the volume of each source region (or cell) of the simulation. As the “true” analytical constructive solid geometry cell volumes are typically not known a priori they must be computed by the application at runtime, which is not straightforward in TRRM as different rays are used each power iteration such that the sampled volume of each cell also changes between iterations. In the present study, we analyze two different on-the-fly stochastic methods for computing the cell volumes and quantify their impacts on the accuracy of scalar flux estimates. We find that the “na¨ıve” stochastic volume estimator (which arises naturally from the derivation of the Method of Characteristics), is highly biased and can result in over 1,000 pcm error in eigenvalue. Conversely, we find that the “simulation averaged” estimator is unbiased and is therefore equivalent to the use of analytical cell volumes even when using a coarse ray density. Thus, the new simulation averaged method is a critical (and as yet undocumented) component of the TRRM algorithm, and is therefore vital information for those in the reactor physics community working to implement random ray solvers of their own.

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“…are 35 made in the process of producing multi-group cross sections for core transport models. In order to avoid such simplifying assumptions, some researchers have recently employed direct full-core Monte Carlo calculations and massive cross section tallies to directly compute multi-40 group cross sections for downstream deterministic transport calculations [2,7]. Such approaches are feasible because Monte Carlo cross section tallies "converge" much more rapidly than local fluxes and power distributions and because cross section tallies can be agglomerated across 45 many spatial regions through use of sophisticated "clustering" or "machine-learning" algorithms [2].…”

Section: Introductionmentioning

confidence: 99%

“…However, none of these methods for generating multigroup cross sections, including Monte Carlo, are capable of directly producing multi-group cross sections that preserve 50 local reaction rates and pin power distributions in 3D core transport models. Even formally "exact" multi-group cross sections lead to errors in downstream core models [2,3] unless additional "equivalence parameters" are introduced to force conservation of local reaction rates.…”

Section: Introductionmentioning

confidence: 99%

Generalized equivalence methods for 3D multi-group neutron transport

Giudicelli

Smith

Forget

2018

Annals of Nuclear Energy

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An analysis of condensation errors in multi-group cross section generation for fine-mesh neutron transport calculations

Cited by 22 publications

References 6 publications

Method of Characteristics for 3D, Full-Core Neutron Transport on Unstructured Mesh

Method of Characteristics for 3D, Full-Core Neutron Transport on Unstructured Mesh

A Comparison of Stochastic Mesh Cell Volume Computation Strategies for the Random Ray Method of Neutral Particle Transport

Generalized equivalence methods for 3D multi-group neutron transport

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