Leakage corrected spatial (assembly) homogenization technique

Rahnema, Farzad; Nichita, Eleodor

doi:10.1016/s0306-4549(96)00084-9

Cited by 40 publications

(17 citation statements)

References 6 publications

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“…the fuel enrichment, which is homogeneously sampled in the interval [2.1%, 3.6%] for both 2. the number of control rods inserted in each heterogeneous fuel bundle (4,8,12,16,24,28), changed keeping the symmetry in the assembly layout…”

Section: Ivb2 Example 2: Uo 2 Colorset With Aic Control Rodsmentioning

confidence: 99%

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Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

Gamarino

Dall’Osso

Lathouwers

et al. 2018

Nuclear Science and Engineering

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-Nodal diffusion is currently the preferred neutronics model for industrial reactor core calculations, which use few-group cross-section libraries generated via standard assembly homogenization. The infinite-medium flux-weighted cross sections fail to capture the spectral effects triggered in the core environment by nonreflective boundary conditions at the fuel-assembly edges. This poses a serious limitation to the numerical simulation of current-and next-generation reactor cores, characterized by strong interassembly heterogeneity.Recently, a spectral rehomogenization method has been developed at AREVA NP. This approach consists of an on-the-fly modal synthesis of the spectrum variation between the environmental and infinite-medium conditions. It uses information coming from both the nodal simulation and the lattice transport calculation performed to compute the standard cross sections. The accuracy of the spectral corrections depends on the choice of the basis and weighting functions for the expansion and on the definition of a realistic energy distribution of the neutron leakage. In this paper, we focus on the first aspect. Two tracks are researched: a combination of analytical functions (with a physically justified mode) and a mathematical approach building upon the Proper Orthogonal Decomposition. The method is applied to relevant pressurized-water-reactor benchmark problems. We show that the accuracy of the cross sections is significantly improved at reasonably low computational cost and memory requirement. Several aspects of the methodology are discussed, such as the interplay with space-dependent corrections. We demonstrate that this approach can model not only the spectral interactions between dissimilar neighbor assemblies but also the spectral effects due to different physical conditions (namely, multiplicative properties) in the environment and in the infinite medium.

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Section: Ivb2 Example 2: Uo 2 Colorset With Aic Control Rodsmentioning

confidence: 99%

“…12 The tabulated corrections are precomputed during the lattice-calculation phase via parametric assembly simulations with varying albedo boundary conditions. Recently, a variant of this approach has been proposed.…”

Section: Introductionmentioning

confidence: 99%

Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

Gamarino

Dall’Osso

Lathouwers

et al. 2018

Nuclear Science and Engineering

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“…Then, the variations of fine-group neutron fluxes cause those of coarsegroup cross sections. In order to capture such physical phenomena, several approaches have been studied (e.g., neutron current to flux ratios on mesh surfaces is considered as a correction index, fine-group neutron fluxes are reconstructed, or several colorset conditions are combined [10][11][12]). In the present study, we consider that the variations of coarse-group cross sections are functions of neutron leakage and propose a new correction technique for coarse-group cross sections considering the spectral interference effect by focusing on the variations of neutron leakage.…”

Section: Overview Of Energy Collapsingmentioning

confidence: 99%

A new technique for spectral interference correction on pin-by-pin BWR core analysis

Fujita

Endo

Yamamoto

2014

Journal of Nuclear Science and Technology

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A new correction technique to capture the spectral interference effect on collapsed cross sections, which focuses on application to the pin-by-pin boiling water reactor (BWR) core analysis, is proposed. The spectral interference effect, which is caused by adjacent loadings of different types of fuel assemblies, has relationship with variations of neutron leakage in each pin-cell from the viewpoint of neutron balance. Variation of neutron leakage affects neutron spectrum and thus the neutron leakage is considered to be important to correct coarse-group cross sections used in core calculations. We focus on the neutron leakage in each pin-cell and use it as a correction index (i.e., a leakage index (LI)), which is defined as the volume-averaged neutron leakage in a pin-cell. By utilizing the leakage index, we represent the variations of coarse-group cross sections as the linear combination of LIs. In order to verify and discuss the applicability of the present correction technique, two-dimensional benchmark calculations considering typical characteristics of BWR cores are carried out. From the calculation results, the present correction technique well reproduces the reference coarse-group cross sections and improves the calculation accuracies.

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“…The one dimensional benchmarks are slabs of assemblies which are simplified representations of Boiling Water Reactor (BWR) configurations, while the two dimensional benchmark problems are two small subsets of the CISE core and the complete CISE core (Rahnema and Nichita, 1997).…”

Section: Numerical Resultsmentioning

confidence: 99%

“…One dimensional BWR problems composed of seven assemblies with heterogeneous pins extracted from Rahnema and Nichita, (1997). The reactor has three different configurations constructed by combining three different assemblies to represent a BWR assembly.…”

Section: B5 One Dimensional Bwr Problemsmentioning

confidence: 99%

Development of a finite element method for neutron transport equation approximations

Ferràndiz¹

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Supervisors: Damián Ginestar Peiró Gumersindo Verdú Martín Ací em pariren i ací estic. I com que he investigat certes coses, ací les conte, ací les dic. AbstractThe neutron transport equation describes the neutron population and the nuclear reactions inside a nuclear reactor core. First, this equation is introduced and its assumptions are stated. Then, the stationary neutron diffusion equation which is the most useful approximation of this equation, is studied. This approximation leads to a differential eigenvalue problem. To solve the neutron diffusion equation, a h ≠ p finite element method is investigated. To improve the efficiency of the method a Restricted Additive Schwarz preconditioner is implemented.Once the solution for the steady state neutron distribution is obtained, it is used as initial condition for the time integration of the neutron diffusion equation. To test the behaviour of the method, rod ejection accidents are numerically simulated. However, a non-physical behaviour appears when a cell is partially rodded: this is, the rod cusping effect, which is solved by using a moving mesh scheme. In other words, the mesh follows the movement of the control rod. Numerical results show that the rod cusping effect is corrected with this scheme.After that, the simplified spherical harmonics approximation, SP N , is developed to solve the steady state problem. This approximation extends the spherical harmonics approximation, P N , in one dimensional geometries to multidimensional geometries with strong assumptions. It improves the diffusion theory results but does not converge as N tends to infinity. The advantages and limitations of this approximation are tested on several one-, two-and three-dimensional reactors.Finally, the spatial homogenization in the context of the finite element method is studied. Homogenization consists in replacing heterogeneous subdomains by homogeneous ones, in such a way that the homogenized problem provides fast and accurate average results. Discontinuous solutions were proposed in the Generalized Equivalence Theory. Here, a discontinuous Galerkin finite element method where the jump condition for the neutron flux is imposed in a weak sense using interior penalty terms is introduced. Also, the use of discontinuity factors for the correction of the homogenization error when using the SP N equations is investigated.v ResumenLa ecuación del transporte neutrónico describe la población de neutrones y las reacciones nucleares dentro de un reactor nuclear. Primero, introducimos esta ecuación y las aproximaciones de la misma. Entonces, estudiamos la ecuación de la difusión neutrónica, la aproximación al transporte más utilizada. Para el caso estacionario, esta aproximación da lugar a un problema diferencial de valores propios. Para resolver la ecuación de la difusión se ha desarrollado un método de elementos finitos h≠p. Para mejorar la eficiencia del método se ha implementado un precondicionador del tipo Restricted Additive Schwarz.Una vez hemos obtenido la distribución neutrónica en ...

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Leakage corrected spatial (assembly) homogenization technique

Cited by 40 publications

References 6 publications

Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

Rehomogenization of Nodal Cross Sections via Modal Synthesis of Neutron Spectrum Changes

A new technique for spectral interference correction on pin-by-pin BWR core analysis

Development of a finite element method for neutron transport equation approximations

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