Simulation of a sodium fast core: Effect of B 1 leakage models on group constant generation

Faure, Bastien; Marleau, G.

doi:10.1016/j.anucene.2016.10.002

Cited by 5 publications

(4 citation statements)

References 8 publications

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“…These tendencies confirm an earlier conclusion of Faure et al [10] whereas leakage-corrected XS generation methods were studied in in fast systems. It was reported that the leakage models had a low impact on cross section generation.…”

Section: Few-group Cross Section Generation Sensitivity Studysupporting

confidence: 89%

Analysis of the Oecd/Nea SFR Benchmark With Ants Reduced-Order Nodal Diffusion Solver and the Serpent Monte Carlo Code

Szogradi

2021

EPJ Web Conf.

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In order to meet modern industrial and scientific demands the Kraken multi-physics platform’s development was recently launched at VTT Technical Research Centre of Finland. The neutronic solver of the framework consists of two calculation chains, providing full core solutions by the Serpent high fidelity code (1) and the AFEN/FENM-based reduced-order diffusion solver called Ants (2) capable of handling square and hexagonal geometries in steady-state. Present work introduces the simulation of a large 3600 MWth Sodium-cooled Fast Reactor (SFR) described within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of OECD. Full-core 3D results were obtained by Serpent for carbide- and oxide-fuel cores, moreover group constants were generated for Ants utilizing 2D super-cell and single assembly infinite lattice models of Serpent. The continuous-energy Monte Carlo method provided the reference results for the verification of the reduced-order method. Implementing the spatially homogenized properties, 3D solutions were obtained by Ants as well for both core configurations. Comparison was made between the various core designs and codes based on reactivity feedbacks (Doppler constant, sodium voiding, control rod worth) considering power distributions. Regarding reactivity sensitivity on geometry, axial fuel- and radial core expansion coefficients were evaluated as well.

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Section: Few-group Cross Section Generation Sensitivity Studysupporting

confidence: 89%

Analysis of the Oecd/Nea SFR Benchmark With Ants Reduced-Order Nodal Diffusion Solver and the Serpent Monte Carlo Code

Szogradi

2021

EPJ Web Conf.

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“…The DRAGON/DONJON calculations were performed using a deterministic reactor physics code package DRAGON version 5, developed at Ecole Polytechnique de-Montr eal [21,22], along with the JEFF-3.1.1 nuclear data. Lattice calculations with 172 groups library were conducted with DRAGON (lattice code) using transport method to generate the homogenized 24-group energy grid [23] cross sections of different subassemblies. The full core calculations of SPX were conducted with DON-JON (full core simulation code) and the homogenized 24-group cross sections, using a 3D hexagonal-z geometry.…”

Section: Serpent/dyn3djjef311mentioning

confidence: 99%

Superphénix Benchmark Part I: Results of Static Neutronics

Ponomarev

Mikityuk

Zhang

et al. 2021

Journal of Nuclear Engineering and Radiation Science

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In the paper, the specification of a new neutronics benchmark for a large Sodium cooled Fast Reactor core and results of modelling by different participants are presented. The neutronics benchmark describes the core of the French sodium cooled reactor Superphénix at its startup configuration, which in particular was used for experimental measurement of reactivity characteristics. The benchmark consists of the detailed heterogeneous core specification for neutronic analysis and results of the reference solution. Different core geometries and thermal conditions from cold “as fabricated” up to full power were considered. The reference Monte Carlo solution of Serpent 2 includes data on multiplication factor, power distribution, axial and radial reaction rates distribution, reactivity coefficients and safety characteristics, control rods worth, kinetic data. The results of modelling with seven other solutions using deterministic and Monte Carlo methods are also presented and compared to the reference solution. The comparisons results demonstrate appropriate agreement of evaluated characteristics. The neutronics results will be used in the second phase of the benchmark for evaluation of transient behaviour of the core.

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“…The present paper focuses on the B 1 method for leakage-corrected group constant generations by the Monte Carlo method. The B 1 method is generally used for representing the neutron leakage in many deterministic calculations (Duderstadt and Hamilton, 1976;Deniz, 1986;Petrovic and Benoist, 1996;Rimpault et al, 2002;Hébert, 2009;van Rooijen and Chiba, 2011;Faure and Marleau, 2017). Implementation of the B method into the Monte Carlo method encounters difficulty in treating complex-valued neutron flux or neutron current.…”

Section: Introductionmentioning

confidence: 99%

“…The capability of calculating direction-dependent diffusion coefficients is desirable because the direction dependence of diffusion coefficients is notable in a fertile fuel assembly (Faure and Marleau, 2017) and a void containing fuel pin cell (Yamamoto, 2012). Previous works on the directional diffusion coefficients by the Monte Carlo method are based on various techniques, and they are published in Milgram (1997), Gelbard and Pego, (1979), Yamamoto (2012), Dorval and Leppänen (2015), Dorval (2016a), and Dorval (2016b).…”

Section: Introductionmentioning

confidence: 99%

Monte Carlo method for solving a B1 equation with complex-valued buckling in asymmetric geometries and generation of directional diffusion coefficients

Yamamoto

Sakamoto²

2018

Annals of Nuclear Energy

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Direction-dependent diffusion coefficients generated using a Monte Carlo method that explicitly solves a heterogeneous B 1 equation are validated in this paper. The validation is conducted by comparing the direction-dependent neutron leakage using directional diffusion coefficients with the results of a neutron transport calculation. This paper develops a new method for critical buckling search calculation to solve a heterogeneous B 1 equation for an asymmetric unit cell in which the critical bucking has a complex value. A complex-valued critical buckling can be searched using the differential operator sampling method for estimating the sensitivity coefficients of k eff with respect to the buckling. In an asymmetric unit cell, the neutron flux and the neutron current also have complex values. The real part of the neutron flux obtained by a critical buckling search calculation represents the neutron spectrum in the center portion of the finite geometry composed of the unit cells.

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Simulation of a sodium fast core: Effect of B 1 leakage models on group constant generation

Cited by 5 publications

References 8 publications

Analysis of the Oecd/Nea SFR Benchmark With Ants Reduced-Order Nodal Diffusion Solver and the Serpent Monte Carlo Code

Analysis of the Oecd/Nea SFR Benchmark With Ants Reduced-Order Nodal Diffusion Solver and the Serpent Monte Carlo Code

Superphénix Benchmark Part I: Results of Static Neutronics

Monte Carlo method for solving a B1 equation with complex-valued buckling in asymmetric geometries and generation of directional diffusion coefficients

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