Reactor physics analysis of the SPERT III E-core with Tripoli-4®

Zoia, Andrea; Brun, Emeric

doi:10.1016/j.anucene.2015.11.032

Cited by 19 publications

(4 citation statements)

References 25 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The core layout of the SPERT-III facility [12] is illustrated in Figure 7, with a colored map for the properties identifying the resonant nuclides used in the self-shielding calculations over the full north-east quadrant. Despite the fact that the fuel pins are all the same, different colors are automatically selected to request different self-shielded cross sections spatially.…”

Section: Examples Of Typical and Advanced Use Casesmentioning

confidence: 99%

“…A radial cut of the Jules Horowitz Reactor (JHR) [11] produced by ALAMOS is illustrated in Figure 6, together with a particular detail of the unit cell, showing broad use of element overlap and automatic mesh recalculation by the method intersector2d. The core layout of the SPERT-III facility [12] is illustrated in Figure 7, with a colored map for the properties identifying the resonant nuclides used in the self-shielding calculations over the full north-east quadrant. Despite the fact that the fuel pins are all the same, different colors are automatically selected to request different self-shielded cross sections spatially.…”

Section: Examples Of Typical and Advanced Use Casesmentioning

confidence: 99%

See 1 more Smart Citation

Overview of SERMA’s Graphical User Interfaces for Lattice Transport Calculations

et al. 2022

View full text Add to dashboard Cite

This article presents an overview of the graphical user interfaces (GUIs) developed at CEA/SERMA (Service d’Études des Réacteurs et de Mathématiques Appliquées) in Saclay, France, which have been used for over forty years by engineers and scientists to build geometries and meshes for general-purpose lattice transport calculations (neutrons and photons). Several applications make use of these calculations, from fuel assembly to full core design, criticality and safety, needing consistency check of the geometry and input properties before starting any lattice calculation. The software pattern design of the GUIs is briefly discussed, showing also the rationale behind the two interfaces for the construction of the geometries for simple fuel assemblies and complex motifs including the reflector (colorsets). The new GUI, ALAMOS, specifically developed for APOLLO3® with a Python Application Programming Interface (API), is here presented as the successor of Silène, which was the first GUI released in the 1990s to serve APOLLO2 calculations. The considerable experience gained by Silène over the years with plenty of various applications has provided a crucial support for the development of ALAMOS.

show abstract

Section: Examples Of Typical and Advanced Use Casesmentioning

confidence: 99%

Section: Examples Of Typical and Advanced Use Casesmentioning

confidence: 99%

Overview of SERMA’s Graphical User Interfaces for Lattice Transport Calculations

et al. 2022

View full text Add to dashboard Cite

show abstract

“…Assessing the asymptotic behaviour of a nuclear system is intimately related to computing the dominant direct (Here and throughout the manuscript, the terms 'direct' and 'forward' are equivalently used.) eigenmode and the associated dominant eigenvalue of the configuration under analysis, which is central in several applications in reactor physics, including pulsed neutron reactivity measurements [1] and reactor period analysis [2][3][4]. Furthermore, for several key reactor coefficients, such as kinetics parameters, sensitivities to nuclear and material data, or perturbations [5][6][7][8][9][10], bilinear forms involving both the direct and the adjoint fundamental eigenfunctions are required.…”

Section: Introductionmentioning

confidence: 99%

Comparison of Direct and Adjoint k and α-Eigenfunctions

Vitali

Chevallier

Jinaphanh

et al. 2021

JNE

Self Cite

View full text Add to dashboard Cite

Modal expansions based on k-eigenvalues and α-eigenvalues are commonly used in order to investigate the reactor behaviour, each with a distinct point of view: the former is related to fission generations, whereas the latter is related to time. Well-known Monte Carlo methods exist to compute the direct k or α fundamental eigenmodes, based on variants of the power iteration. The possibility of computing adjoint eigenfunctions in continuous-energy transport has been recently implemented and tested in the development version of TRIPOLI-4®, using a modified version of the Iterated Fission Probability (IFP) method for the adjoint α calculation. In this work we present a preliminary comparison of direct and adjoint k and α eigenmodes by Monte Carlo methods, for small deviations from criticality. When the reactor is exactly critical, i.e., for k0 = 1 or equivalently α0 = 0, the fundamental modes of both eigenfunction bases coincide, as expected on physical grounds. However, for non-critical systems the fundamental k and α eigenmodes show significant discrepancies.

show abstract

“…The comprehensive experimental data of the SPERT III E-core program provides a good opportunity to assess the ability of reactor kinetic codes to predict the reactor dynamic behaviour in the event of a reactivity initiated accident (RIA). In the last couple of years, analyses of the static core tests as well as of the reactivity accident tests have already been performed at several research institutes for code validation [3][4][5][6][7][8].…”

Section: Introductionmentioning

confidence: 99%

Transient Calculations of Spert Iii Experiments

Pautz

Zwermann

2021

EPJ Web Conf.

View full text Add to dashboard Cite

Cold-startup and hot-standby reactivity accident tests conducted at the SPERT III E-core research reactor are analysed with the coupled neutron-kinetic/thermal-hydraulic code system DYN3D-ATHLET. Homogenised 2-group cross sections for DYN3D are thereby generated with the Monte Carlo neutron transport code Serpent 2 in combination with the ENDF/B-VII.1 cross section library. Results in terms of maximum power, energy release, and reactivity compensation are in good agreement with the experimental values. The time-dependent contributions to the reactivity feedback are investigated for both a cold-startup test and a hot-standby test. These findings prove the suitability of the combined application of the simulation codes to predict the reactor dynamic behaviour in the event of prompt-critical and super-prompt critical transients even for small reactor cores. Furthermore, static core characteristics of the SPERT III E-core reactor at cold-startup condition are analysed with using a static DYN3D model, a detailed Serpent reference model, and a simplified Serpent model consistent with the DYN3D model. The critical control rod position and the excess reactivities of both the control rods and the transient rod obtained with the Serpent reference model are consistent with the experimental values. For the same parameters, the DYN3D model is in good agreement with the Serpent simplified model.

show abstract

Reactor physics analysis of the SPERT III E-core with Tripoli-4®

Cited by 19 publications

References 25 publications

Overview of SERMA’s Graphical User Interfaces for Lattice Transport Calculations

Overview of SERMA’s Graphical User Interfaces for Lattice Transport Calculations

Comparison of Direct and Adjoint k and α-Eigenfunctions

Transient Calculations of Spert Iii Experiments

Contact Info

Product

Resources

About