Extension of Askew's Coarse Mesh Method to Few-Group Problems for Calculating Two-Dimensional Power Distribution in Fast Breeder Reactors

Takeda, Toshikazu; Komano, Yasuo

doi:10.3327/jnst.15.523

Cited by 13 publications

(2 citation statements)

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“…The presence of the FFCP module makes it possible to calculate a heterogeneous core structure. A multigroup, three-dimensional diffusion problem is discretized by means of the improved coarse-mesh method [9], which is a modification of the method of [10], first proposed for single-group problems and then extended to the multigroup case [11]. The modification relies on a more accurate model to describe the group leakage of neutrons through each lateral boundary of an elementary hexagonal prism.…”

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Accuracy analysis of BN-600 power density calculations

et al. 2010

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Methodological program techniques for performing neutron-physical calculations of fast-reactor cores with a description of the main codes TRIGEX, JARFR, GEFEST, MMKKENO, and ModExSys are described. The results of verification and assessment of the methodological errors on the basis of a test model of BN-600 are presented. Transport effects and mesh errors in the reaction rate distributions in the core, lateral screen, and in-reactor storage area are evaluated. An integral assessment of the computational accuracy is performed by comparing with the experimental data obtained by γ-scans of BN-600 fuel assemblies. It is shown that calculations describe the experimental data over the core with an error no worse than 5%. In the lateral screen and the in-reactor storage area, the discrepancy between the calculations and experiment does not exceed 20-30%.Computational modeling of the neutron field is the main method of determining the energy release in a fast-reactor core, where, as a rule, there is no apparatus for tracking the power density. Correspondingly, computational accuracy is one of the key elements of the operational reliability of fuel elements and assemblies. Another aspect of the problem is substantiating the conservatism of fast-reactor core design -design limits or margins affecting the technical-economic characteristics. Experience in operating BN-600 makes it possible to evaluate the accuracy of the design and operational programs under real conditions of a commercial medium-power fast reactor and to take it into account in subsequent development work.Such an analysis is based on the measurement of the power density in BN-600 by means of γ-scanning. Eleven such experiments were completed in the time period from physical startup of the reactor in 1980 and up to 2008. The present article analyzes the last experiments performed in 2003-2006 in the course of transitioning from a 01M1 core to a 01M2 core with maximum burnup 11.1% [1]. This upgrade was an important next step in making use of the serviceability margin of reactor fuel assemblies, which required greater attention to monitoring the characteristics of the core [2].Program and Constants System for Neutron-Physical Calculations of Fast Reactors. A consistent system for performing neutron-physical calculations of fast reactors is now available. It includes the program systems TRIGEX [3], JARFR [4], GEFEST [5], and MMKKENO [6] as well as the BNAB constants library [7,8]. It should be noted that the base

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Accuracy analysis of BN-600 power density calculations

et al. 2010

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“…SHARP is a two dimensional, two group diffusion theory code with the capability to take into account feedback effects caused by thermal-hydraulics and fission products, to execute a boron concentration search during burnup calcu-lation, and to perform a pin power reconstruction (16). In the SHARP calculations, the fuel assemblies are divided into 2x2 node, and the improved coarse mesh method (17) or the advanced nodal method(18) can be utilized to reduce spatial mesh effect.…”

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Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors.

Yamamoto

Kanda

1997

J. Nucl. Sci. Technol.

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Analyses of an equilibrium cycle are useful for evaluating newly designed fuels, defining an envelope of core operating parameters, and so on. However, generation of a loading pattern for the equilibrium cycle is much more difficult than that of a single cycle. Therefore, a loading pattern optimization code for the equilibrium cycle of pressurized water reactors, OPAL, has been newly developed on the basis of the simulated annealing method. In order to verify the capability of the OPAL code, comparison with successive multicycle optimizations was performed while fixing the number of fresh fuel in each cycle. Through benchmark calculations, it was found that the result of the equilibrium cycle optimization was almost compatible with that of the successive multicycle optimization, when the definition of each objective function was similar. However, successive multicycle optimization includes some ambiguity due to limits on the number of calculated cycles, since it requires much computation time. Consequently, the equilibrium cycle optimization has advantages including the quantitative comparison of the core neutronic performances.

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Analytic derivation of the correction factor for the improved coarse mesh method

Yamamoto¹

2004

Annals of Nuclear Energy

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Extension of Askew's Coarse Mesh Method to Few-Group Problems for Calculating Two-Dimensional Power Distribution in Fast Breeder Reactors

Cited by 13 publications

References 2 publications

Accuracy analysis of BN-600 power density calculations

Accuracy analysis of BN-600 power density calculations

Comparison between Equilibrium Cycle and Successive Multicycle Optimization Methods for In-Core Fuel Management of Pressurized Water Reactors.

Analytic derivation of the correction factor for the improved coarse mesh method

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