Reconstruction of Pin Power and Burnup Characteristics from Nodal Calculations in Hexagonal-Z Geometry

Yang, Won Sik; Finck, P.J.; Khalil, Hassan A.

doi:10.13182/nse92-a23920

Cited by 18 publications

(8 citation statements)

References 5 publications

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“…As an example of the accuracy of the overall computational schemes, Table 15 shows the calculation-to-experiment (C/E) comparisons for the fuel pin burnups of an Integral Fast Reactor (IFR) [78] test assembly irradiated in EBR-II. Fuel pin burnups were computed at several different axial positions using the DIF3D-Nodal flux solution option and the RCT code [79] for reconstruction of pin power and burnup characteristics.…”

Section: Depletion and Fuel Cycle Analysis Codesmentioning

confidence: 99%

Fast Reactor Physics and Computational Methods

Yang¹

2012

Nuclear Engineering and Technology

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Section: Depletion and Fuel Cycle Analysis Codesmentioning

confidence: 99%

Fast Reactor Physics and Computational Methods

Yang¹

2012

Nuclear Engineering and Technology

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“…4 Instead of using an arithmetic average for the diffusion coefficients, flux-weighted or possibly current-weighted averages could also be used.…”

Section: Interior Node Corner Point Approximationmentioning

confidence: 99%

“…Since then developments in flux reconstruction techniques for depletion calculations in hexagonal geometry have provided acceptable approximations to the corner flux values based on surface-and node-averaged flux values. 4,5 These corner point flux values, along with surface-averaged flux values, can then be used to approximate the flux shape on the surfaces of a node. The new method presented in this report attempts to accurately capture the neutron balance by including the previously neglected discontinuous terms.…”

Section: Introductionmentioning

confidence: 99%

Nodal Green?s Function Method Singular Source Term and Burnable Poison Treatment in Hexagonal Geometry

Bingham¹,

Ferrer²,

Ougouag³

2009

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An accurate and computationally efficient two-or three-dimensional (2-D or 3-D) neutron diffusion model will be needed to develop, compute safety parameters, and analyze the fuel cycle of a prismatic very high temperature reactor design under the Next Generation Nuclear Plant. To meet this need, an analytical Nodal Green's function method for the transverse integrated neutron diffusion equation was developed in 2-D and 3-D hexagonal geometry. This scheme is incorporated into the algorithm of the HEXPEDITE, a code first developed by Fitzpatrick and Ougouag. HEXPEDITE neglects discontinuous terms that arise in the transverse leakage because of the transverse integration procedure application to hexagonal geometry, and cannot account for the localized effects of burnable poisons across nodal boundaries. The test code being developed for this document accounts for these terms by maintaining a strict inventory of neutrons by using the nodal balance equation as a constraint of the neutron flux equation. The method developed in this report is intended to restore neutron conservation and increase the accuracy and fidelity of the code by adding the effect of the discontinuous terms to the transverse integrated flux solution. This is achieved through the use of the Nodal Green's function method to the complete equation resulting from the transverse integration procedure. The result is a rigorous semi-analytical solution. The new treatment of the burnable applies a similar approach, but with singular terms of pre-computed strength, based on blackness theory. vi vii ACKNOWLEDGEMENTS

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“…The RCT code [15] developed at Argonne has a capability for calculating homogenous fluxes and power shapes within a fuel assembly using the assembly-average and surface-average data obtained from DIF3D nodal diffusion calculations. Since it was developed for fast reactor application, however, it does not use pin power form functions.…”

Section: Pin Power Factorsmentioning

confidence: 99%