Building neutron cross-section dependencies for few-group reactor calculations using stepwise regression

Shchukin, N. V.; Semenov, Andrey A.

doi:10.1016/j.anucene.2004.06.009

Cited by 27 publications

(7 citation statements)

References 4 publications

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“…135 Xe and 135 I are needed for instance to study the power fluctuations due to shortterm control. About PWR, 10 B is included because it is responsible for long-term control. A residual material lumps all other nuclides which are not considered by the depletion solver of the core code.…”

Section: Specifications Of the Pin-by-pin Data Modelmentioning

confidence: 99%

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A multivariate representation of compressed pin-by-pin cross sections

Tomatis

2021

EPJ Nuclear Sci. Technol.

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Since the 80’s, industrial core calculations employ the two-step scheme based on prior cross sections preparation with few energy groups and in homogenized reference geometries. Spatial homogenization in the fuel assembly quarters is the most frequent calculation option nowadays, relying on efficient nodal solvers using a coarse mesh. Pin-wise reaction rates are then reconstructed by dehomogenization techniques. The future trend of core calculations is moving however toward pin-by-pin explicit representations, where few-group cross sections are homogenized in the single pins at many physical conditions and many nuclides are selected for the simplified depletion chains. The resulting data model requires a considerable memory occupation on disk-files and the time needed to evaluate all data exceeds the limits for practical feasibility of multi-physics reactor calculations. In this work, we study the compression of pin-by-pin homogenized cross sections by the Hotelling transform in typical PWR fuel assemblies. The reconstruction of these quantities at different physical states of the assembly is then addressed by interpolation of only a few compressed coefficients, instead of interpolating separately each homogenized cross section. Savings in memory higher than 90% are observed, which result in important gains in runtime when interpolating the few-group data.

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Section: Specifications Of the Pin-by-pin Data Modelmentioning

confidence: 99%

“…The reconstruction of the few-group cross section data as continuous functions over P is usually performed by a linear interpolation scheme [10][11][12]. Here, cross sections are always considered as homogeneous data in each cell of the spatial mesh used for the calculation.…”

Section: Multivariate Polynomial Interpolationmentioning

confidence: 99%

A multivariate representation of compressed pin-by-pin cross sections

Tomatis

2021

EPJ Nuclear Sci. Technol.

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“…(2) TLLI Method The microscopic group cross section evaluation by using an interpolation table is proper for many cases in reactor physics calculation, and many methods have been proposed. 6) In the TLLI method, the linear interpolation is employed from the simplicity of calculation.…”

Section: Modification For Improvement Of Microscopic Groupmentioning

confidence: 99%

Effects of Fuel and Coolant Temperatures and Neutron Fluence on CANDLE Burnup Calculation

Sekimoto

Udagawa

2006

Journal of Nuclear Science and Technology

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A new method is developed to analyze CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup, where the microscopic group cross-sections are evaluated at every space mesh by TLLI (Table Look-up and Linear Interpolation) method, and used to analyze a fast reactor with natural uranium as a fresh fuel. The results are compared with the conventional method, where only one set of the microscopic group cross-sections is employed, to investigate the effects of space-dependency of the microscopic group cross-sections and feasibility of the old method. The differences of the effective neutron multiplication factor, burning region moving speed, spent fuel burnup and spatial distributions of nuclide densities, neutron fluence and power density may be considerable from the reactor designer point. However, they are small enough when we study about the characteristics of CANDLE burnup for different designs.

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“…Ref. [16] presents a parametric analysis of homogenized cross-sections across operational ranges of burnup, coolant density, coolant temperature, fuel temperature, and boron concentration for a VVER. In addition, a polynomial approximation method using stepwise regression for HXS is demonstrated.…”

Section: Introductionmentioning

confidence: 99%

Methodology for Sensitivity Analysis of Homogenized Cross-Sections to Instantaneous and Historical Lattice Conditions with Application to AP1000® PWR Lattice

et al. 2021

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In the two-step method for nuclear reactor simulation, lattice physics calculations are performed to compute homogenized cross-sections for a variety of burnups and lattice configurations. A nodal code is then used to perform full-core analysis using the pre-calculated homogenized cross-sections. One source of uncertainty introduced in this method is that the lattice configuration or depletion conditions typically do not match a pre-calculated one from the lattice physics simulations. Therefore, some interpolation model must be used to estimate the homogenized cross-sections in the nodal code. This current study provides a methodology for sensitivity analysis to quantify the impact of state variables on the homogenized cross-sections. This methodology also allows for analyses of the historical effect that the state variables have on homogenized cross-sections. An application of this methodology on a lattice for the Westinghouse AP1000® reactor is presented where coolant density, fuel temperature, soluble boron concentration, and control rod insertion are the state variables of interest. The effects of considering the instantaneous values of the state variables, historical values of the state variables, and burnup-averaged values of the state variables are analyzed. Using these methods, it was found that a linear model that only considers the instantaneous and burnup-averaged values of state variables can fail to capture some variations in the homogenized cross-sections.

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Building neutron cross-section dependencies for few-group reactor calculations using stepwise regression

Cited by 27 publications

References 4 publications

A multivariate representation of compressed pin-by-pin cross sections

A multivariate representation of compressed pin-by-pin cross sections

Effects of Fuel and Coolant Temperatures and Neutron Fluence on CANDLE Burnup Calculation

Methodology for Sensitivity Analysis of Homogenized Cross-Sections to Instantaneous and Historical Lattice Conditions with Application to AP1000® PWR Lattice

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