The Serpent Monte Carlo Code: Status, Development and Applications in 2013

Leppänen, Jaakko; Pusa, Maria; Viitanen, Tuomas; Valtavirta, Ville; Kaltiaisenaho, Toni

doi:10.1051/snamc/201406021

Cited by 50 publications

(23 citation statements)

References 58 publications

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“…All neutronic results presented are simulated through a Monte Carlo simulation routine using the Serpent 2.1.27 [23], using data libraries JEFF 3.11 as shown in Figure 4. The simulations are performed using a 100k neutron population with 25 inactive cycles and 25 active cycles to allow for suitable flux convergence.…”

Section: Methodsmentioning

confidence: 99%

Variable Reactivity Control in Small Modular High Temperature Reactors Using Moderation Manipulation Techniques

2018

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Abstract:With extensive research being undertaken into small modular reactor design concepts, this has brought new challenges to the industry. One key challenge is to be able to compete with large scale nuclear power plants economically. In this article, a novel approach is applied to reduce the overall dependence on fixed burnable poisons during high reactivity periods within a high temperature graphite moderated reactor. To reduce the excess activity, we aim to harden the flux spectrum across the core by removing part of the central moderation column, thus breeding more plutonium, in a later period the flux spectrum is softened again to utilise this plutonium again. This provides a neutron storage effect within the 238 U and the resulting breeding of Plutonium. Due to the small size and the annular design of the high temperature reactor, the central reflector is key to the thermalization process. By removing a large proportion of the central reflector, the fuel within the proximity of the central reflector are less likely to receive neutrons within the thermal energy range. In addition to this, the fuel at the extremities of the core have a higher chance of fission due to the higher number of neutrons reaching them. This works as a method of balancing the power distribution between the central and outside fuel pins. During points of low reactivity, such as the end of the fuel cycle, the central reflector can be reinserted and the additionally bred plutonium and U 235 at the centre of the core will encounter a higher probability of fission due to more thermal neutrons within this region. By removing the central reflector, this provided a 320 pcm reactivity drop for the duration of the fuel cycle. The plutonium buildup provided additional fissile material up until the central reflector was reinserted. The described method created a two-fold benefit. The overall full power days within the core was increased bỹ 31 days due to the additional fissile material within the core and secondly the highest loaded power pins saw a 30% power reduction during the removal of the central reflector column.

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Section: Methodsmentioning

confidence: 99%

Variable Reactivity Control in Small Modular High Temperature Reactors Using Moderation Manipulation Techniques

2018

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“…The two-group homogenization parameters are computed by the SERPENT continuousenergy Monte Carlo neutron transport code. 33 Version 2.1.28 of SERPENT is used in combination with the JEFF3.1 nuclear data library. 34 The single-assembly calculations for group-constant generation are run with 750 active cycles of 7.5 × 10 5 source neutrons (50 inactive cycles are discarded to allow the initial fission-source distribution to converge).…”

Section: Iva Proceduresmentioning

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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-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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“…The TFM matrices are calculated with a modified version of Serpent 2.1.21 [7]. The modifications concern the calculation of the fission matrices, including the distinction between prompt and delayed neutrons, the fission to fission time matrix, and the correlated sampling technique to generate the locally perturbed matrices.…”

Section: Calculation Parameters 341 Monte Carlo -Serpentmentioning

confidence: 99%

Towards spatial kinetics in a low void effect sodium fast reactor: core analysis and validation of the TFM neutronic approach

Laureau

Buiron

Fontaine

2017

EPJ Nuclear Sci. Technol.

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Abstract. The studies presented in this paper are performed in the general framework of transient coupled calculations with accurate neutron kinetics models able to characterize spatial decoupling in the core. An innovative fission matrix interpolation model has been developed with a correlated sampling technique associated to the Transient Fission Matrix (TFM) approach. This paper presents a validation of this Monte Carlo based kinetic approach on sodium fast reactors. An application case representative of an assembly of the low void effect sodium fast reactor ASTRID is used to study the physics of this kind of system and to illustrate the capabilities provided by this approach. To validate the interpolation model developed, different comparisons have been performed with direct Monte Carlo and ERANOS deterministic S N calculations on spatial kinetics parameters (flux redistribution, reactivity estimation, etc.) together with point kinetics feedback estimations.

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The Serpent Monte Carlo Code: Status, Development and Applications in 2013

Cited by 50 publications

References 58 publications

Variable Reactivity Control in Small Modular High Temperature Reactors Using Moderation Manipulation Techniques

Variable Reactivity Control in Small Modular High Temperature Reactors Using Moderation Manipulation Techniques

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

Towards spatial kinetics in a low void effect sodium fast reactor: core analysis and validation of the TFM neutronic approach

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