Overview of the NURESAFE European Project

Chanaron, Bruno

doi:10.1016/j.nucengdes.2017.09.001

Cited by 18 publications

(6 citation statements)

References 15 publications

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“…The results in Figures 6,8,9 and 13 show that despite the high power peaking factors in the hottest assemblies only subcooled boiling and small vapor volume fraction in the upper part of the assemblies is predicted, because of the low inlet temperatures.…”

Section: Pepm'2021mentioning

confidence: 95%

“…COBAYA4 [3,4] is a 3D multi-scale core physics code using transport-corrected multigroup diffusion approximation. It is developed by the Universidad Politecnica de Madrid and benchmarked for VVER-1000 calculations [5,6] in the frame of the EU NURISP [7] and NURESAFE [8] projects. At the nodal level the analytical coarse-mesh finitedifference (CMFD) method [9] is used.…”

Section: Full-core Nodal Calculationmentioning

confidence: 99%

“…Use of a realistic VVER-1000 multi-parameter two-group cross-section library [15] for reactivity accident analysis which has been generated with the APOLLO2 code [16] and validated in the frame of the NURESAFE project [8]; Use of time-dependent assembly-by-assembly MSLB thermal-hydraulic core boundary conditions (inlet temperatures, inlet mass flow rates, and outlet pressures) obtained from a full plant simulation involving a quasi-3D reactor pressure vessel TH model [17]; Coarse-mesh CTF thermal-hydraulic model with one channel per assembly and 30 axial nodes in the heated part; Fuel model with 9 radial rings in the fuel, one for the gas gap and one for the cladding; Temperature-dependent fuel and cladding thermal-physical properties [18]; The spacer grids are not explicitly modeled and are taken into account by the vertical pressure loss coefficients;…”

Section: Full-core Nodal Calculationmentioning

confidence: 99%

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Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

2021

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The objective of this paper is to analyze the ability of a VVER-1000 core and its control system to cope with a hypothetical main steam line break (MSLB) accident in case of multiple equipment failures. The study involves the use of advanced 3D core calculation models benchmarked and validated for reactivity accidents in preceding studies. A MSLB core boundary condition problem is solved on a coarse (nodal) mesh with the coupled COBAYA/CTF neutronic/thermal hydraulic codes. The core thermal-hydraulic boundary conditions are obtained from a preceding full-plant MSLB simulation. The assessment of the core safety parameters is supplemented by a fine-mesh (sub-channel) thermal-hydraulic analysis of the hottest assemblies with the CTF code using information from the 3D nodal COBAYA/CTF calculations. Thirteen variants of a pessimistic MSLB scenario are considered, each of them assuming a number of equipment failures aggravated by eight control rods stuck out of the core after scram at different locations in the overcooled sector. The results (within the limitations of the adopted modeling assumptions) show that the core safety parameters do not exceed the safety limits in the simulated aggravated reactivity accidents.

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Section: Pepm'2021mentioning

confidence: 95%

Section: Full-core Nodal Calculationmentioning

confidence: 99%

Section: Full-core Nodal Calculationmentioning

confidence: 99%

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Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

2021

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“…The developed numerical platform is based on the open-source SALOME platform [4], using an approach similar to the NURESAFE platform, developed by the CEA [5]. We use the open-software SALOME platform to add new codes and develop coupling interfaces compatible with open and closed source codes.…”

Section: Introductionmentioning

confidence: 99%

FEMuS-Platform: a numerical platform for multiscale and multiphysics code coupling

Chierici¹,

Barbi²,

Bornia³

et al. 2021

9th Edition of the International Conference on Computational Methods for Coupled Problems in Science and Engineering

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Nowadays, many open-source numerical codes are available to solve physical problems in structural mechanics, fluid flow, heat transfer, and neutron diffusion. However, even if these codes are often highly specialized in the numerical simulation of a particular type of physics, none of them allows simulating complex systems involving all the physical problems mentioned above. In this work we present a numerical framework, based on the SALOME platform, developed to perform multiscale and multiphysics simulations involving all the mentioned physical problems. In particular, the developed numerical platform includes the multigrid finite element in-house code FEMuS for heat transfer, fluid flow, turbulence and fluid-structure modeling; the open-source finite volume CFD software OpenFOAM; the multiscale neutronic code DONJON-DRAGON; and a system-scale code used for thermal-hydraulic simulations. Efficient data exchange among these codes is performed within computer memory by using the MED libraries, provided by the SALOME platform.

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“…• Accurate representation of physical phenomena in reactor and core physics, two-phase flow thermal hydraulics and fuel modeling • Multi-scale and multi-physics capabilities of reactor safety computations Figure 2.8 shows all the codes integrated into SALOME platform and coupling possibilities to simulate the different physical phenomena. All the codes except TRIPOLI4 and CRONOS2 are integrated into NURESIM platform (Chanaron 2017).…”

Section: Nuresimmentioning

confidence: 99%

Study and lmplementation of Neutronic-Thermohydraulic Coupling Methods for Detailed Analysis of Transients in Nuclear Reactors for Safety Margins Optimization

Alcaráz¹

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Understanding the performance of a nuclear reactor is a complex task that covers different physical fields as the neutron behavior or fluid dynamics. These fields are not isolated, and there is a feedback among them. Along nuclear engineering history, different codes have been developed to analyze each of the physics in which the problem can be split in an isolated way. An example of this are the neutronic or the thermal-hydraulic codes. To study the performance of the reactor in a realistic way, it is necessary to take into account the interaction among the physics, or in other words, the codes have to be coupled. xiii ha facilitado en todo lo posible que pudiera seguir con mis estudios. Además, mis compañeros, Jorge, Paloma, Alfonso, Rosario, Carmen, Lustolde, David y Ruben se han asegurado de que acabara y me han hecho pasar buenos ratos, con unos buenos partidos y con frases como: "le vamos a decir a Edith que te quite el wifi" o "mucho ves tú Juego de Tronos".

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Overview of the NURESAFE European Project

Cited by 18 publications

References 15 publications

Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

Thermal-hydraulic analysis of a VVER-1000 core in MSLB conditions

FEMuS-Platform: a numerical platform for multiscale and multiphysics code coupling

Study and lmplementation of Neutronic-Thermohydraulic Coupling Methods for Detailed Analysis of Transients in Nuclear Reactors for Safety Margins Optimization

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