Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

Lázaro, A. M.; Schikorr, M.; Mikityuk, Konstantin; Luca, Ammirabile; Bandini, G.; Darmet, G.; Schmitt, Didier; Dufour, P; Tosello, A.; Gallego, Eduardo; Varas, Gonzalo Jiménez; Bubelis, E.; Ponomarev, A. N.; Kruessmann, R.; Struwe, D.; Stempniewicz, M.M.

doi:10.1016/j.nucengdes.2014.02.029

Cited by 14 publications

(3 citation statements)

References 4 publications

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“…The TRACE code has been applied for analysis of different SFR designs in the past, see e.g. Lázaro et al, 2014 andBubelis et al, 2017. The neutronics parameters used for the transient calculations were obtained with Serpent 2, 3D continuousenergy Monte Carlo neutron transport code for reactor physics application being developed in VTT Technical Research Centre of Finland since 2004 by Leppänen et al, 2015. The JEFF311 nuclear data library was used in these calculations.…”

Section: Codes Used For Simulationsmentioning

confidence: 99%

Analysis of Hypothetical Unprotected Loss of Flow in Superphénix Start-Up Core: Sensitivity to Modeling Details

Ponomarev

Mikityuk

2019

ICONE

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The Superphénix (SPX) Sodium Fast Reactor (SFR) core behaviour in a hypothetical Unprotected Loss Of Flow (ULOF) transient scenario is simulated and the impact of the core modelling options on the results are assessed. The analysis has been performed using the TRACE system code modified for the fast reactor applications (Mikityuk et al., 2005) using a point kinetics neutronics model and 1D multichannel thermal-hydraulic model. The SPX start-up core benchmark specification (Ponomarev, et al., 2018) was used for calculation of neutron physics data employed as input in the TRACE simulations. The analysis covered transient simulation before sodium boiling onset. The influence of control rod drive lines (CRDL) expansion reactivity effect on the grace time before boiling was found to be significant. Series of calculations have been performed in order to assess sensitivity of the results to the core specification data and modelling details in both thermal hydraulics and neutronics inputs. In particular, two different models representing fuel subassemblies (three-channel and multichannel) have been employed. In addition, the influence of reactivity effects uncertainty on transient behaviour has been assessed, as well as sensitivity of the results to account for the spatial distribution of reactivity effects in the core. The results of this study serve further as a basis for the transient benchmark exercise proposed by Ponomarev, et al.

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Section: Codes Used For Simulationsmentioning

confidence: 99%

Analysis of Hypothetical Unprotected Loss of Flow in Superphénix Start-Up Core: Sensitivity to Modeling Details

Ponomarev

Mikityuk

2019

ICONE

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“…Once steady-state was verified, a transient [3] consisting of the mass flow reduction of PPs were performed, in both models. Table 2.…”

Section: Simulationsmentioning

confidence: 99%

Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

Ródenas

Chueca

Alsina

2016

EPJ Nuclear Sci. Technol.

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Abstract. One of the goals of Generation IV reactors is to increase safety from those of previous generations. Different research platforms have been identified the need to improve the reliability of the simulation tools to ensure the capability of the plant to accommodate the design basis transients established in preliminary safety studies. The paper describes the modelling of primary pumps in advanced sodium cooled reactors using the TRACE code. Following the implementation of the models, the results obtained in the analysis of different design basis transients are compared with the simplifying approximations used in reference models. The paper shows the process to obtain a consistent pump model of the ESFR (European Sodium Fast Reactor) design and the analysis of loss of flow transients triggered by pumps coast-down analyzing the thermal hydraulic neutronic coupled system response. A sensitivity analysis of the system pressure drops effect and the other relevant parameters that influence the natural convection after the pumps coast-down is also included.

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“…Primary three-dimensional nodalisation -ESFRThe core is represented by 14 axial levels and 4 radial sectors. These radial sectors are representing the inner fuel (1) outer fuel(2) and radial reflector (1) assemblies. 10 of these axial levels and 4 radial rings are representing the active core and axial reflector, so are coupled with 12 Heat Structure components (4 radial sectors x 3 azimuthal sectors) modelling the fuel and reflector assemblies.…”

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confidence: 99%

Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors.

Chueca¹

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Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

Cited by 14 publications

References 4 publications

Analysis of Hypothetical Unprotected Loss of Flow in Superphénix Start-Up Core: Sensitivity to Modeling Details

Analysis of Hypothetical Unprotected Loss of Flow in Superphénix Start-Up Core: Sensitivity to Modeling Details

Pumps modelling of a sodium fast reactor design and analysis of hydrodynamic behavior

Development, assessment and application of computational tools for design safety analysis of liquid metal cooled fast breeder reactors.

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