Boulogne-Billancourt, France
Boulogne-Billancourt, France

Time filter

Source Type

Peniguel C.,Électricité de France | Rupp I.,Électricité de France | Rolfo S.,University of Manchester | Guillaud M.,INCKA
International Congress on Advances in Nuclear Power Plants 2010, ICAPP 2010 | Year: 2010

Fast reactors with liquid metal coolant have recently received a renewed interest owing to a more efficient usage of the primary uranium resources, and they are one of the proposals for the next Generation IV. In order to evaluate nuclear power plant design and safety, 3D analysis of the flow and heat transfer in a wire spacer fuel assembly are ongoing at EDF. The introduction of the wire wrapped spacers, helically wound along the pin axis, enhances the mixing of the coolant between sub-channels and prevents contact between the fuel pins. The mesh generation step constitutes a challenging task if a reasonable amount of cells in conjunction with a suitable spatial discretization is wanted, especially if industrial cases with up to 271 pins need to be tackled as shown in this paper. Complex global flow patterns are found using either k-ε or preferably Reynolds Stress turbulent models with a strong influence of the number of pins. Global parameters like friction factor or Nusselt number are compared against experimental correlations. Likewise exploratory conjugated heat transfer calculations using a coupling between the open source finite element thermal code SYRTHES and open source the finite volume CFD code Code-Saturne, both developed at EDF, are also shown.

Rolfo S.,University of Manchester | Peniguel C.,Électricité de France | Guillaud M.,INCKA | Laurence D.,University of Manchester | Laurence D.,Électricité de France
Nuclear Engineering and Design | Year: 2012

The paper presents refined three-dimensional simulations of the flow and heat transfer in fuel assemblies as found or suggested for liquid metal coolant fast reactors. The wire spacers, helically wound along the pin axis, generate a strong secondary flow pattern in opposition to smooth pins. The eddy viscosity and second moment turbulence models yield to very similar predictions of global friction and heat transfer coefficients and within the range of available experimental correlations. The four configurations simulated range from a small test rig to the full scale reactor bundle (7, 19, 61 and 271 pins) in order to separate (a) global swirl boundary effects, where helical wires leaning against the casing deflect the flow in unison, from (b) homogeneous flow patterns in the core, where wire helices counteract each other. The 61 and 271 pins simulations show a clear decoupling of (a) and (b). The effect of the variation of the helix pass has also been investigated for the 7 pin geometry. Finally, the paper considers the variation of the results as function of the meshing and in particular with the level of detail used for the pin-wire connection. © 2011 Elsevier B.V. All rights reserved.

Li-Puma A.,CEA Saclay Nuclear Research Center | Jaboulay J.-C.,CEA Saclay Nuclear Research Center | Martin B.,Incka
Fusion Engineering and Design | Year: 2014

SYCOMORE, a fusion reactor system code based on a modular approach is under development at CEA. Within this framework, this paper describes the relevant sub-modules which have been implemented to model the main outputs of the breeding blanket and shield block of the system code: tritium breeding ratio, peak energy deposition in toroidal field coils, reactor layout and power deposition, blanket pressure drops and materials inventory. Blanket and shield requirements are calculated by several sub-modules: the blanket assembly and layout sub-module, the neutronic sub-module, the blanket design sub-module (thermal hydraulic and thermo-mechanic pre-design tool). A power flow module has also been developed which is directly linked to the blanket thermo-dynamic performances, which is not described in this paper. For the blanket assembly and layout and the blanket module design sub-modules, explicit analytic models have been developed and implemented; for the neutronic sub-module neural networks that replicate the results of appropriate simplified 1D and 2D neutronic simulations have been built. Presently, relevant model for the Helium Cooled Lithium Lead is available. Sub-modules have been built in a way that they can run separately or coupled into the breeding blanket and shield module in order to be integrated in SYCOMORE. In the paper, the objective and main input/output parameters of each sub-module are reported and relevant models discussed. The application to previous studied reactor models (PPCS model AB, DEMO-HCLL 2006-2007 studies) is also presented. © 2014 Elsevier B.V.

Peniguel C.,Électricité de France | Rupp I.,Électricité de France | Leroyer H.,Électricité de France | Guillaud M.,INCKA
International Conference on Nuclear Engineering, Proceedings, ICONE | Year: 2012

Within the framework of the 2006 French law on sustainable management of radioactive materials and waste, an evaluation of the industrial perspectives of GENERATION iV Fast Reactors deployment is requested for 2012. in this context, questions about waste storage capacities are of interest for EDF (Electricité de France). The storage area is driven by thermal requirements on the argilite formation surrounding the canisters. The size of the final radioactive waste geological repository is one indicator of interest in these studies. The repository considered has been proposed by ANDRA, the French national agency in charge of the radioactive waste management. Optimisation studies on the number of canisters per cell or the distance between two cells using the open source thermal code SYRTHES and SALOME have been presented in previous conferences. However, due to the very large number of calculations induced they are performed on a limited domain. To evaluate the thermal conservatism of a reduced domain, three levels of modeling have been used: a domain limited to half a cell, a slice of half a module (ie three half cells), and a complete half module leading to a very large finite element model (more than 82 million cells), in which all the disposal cells (around 150 of them containing 6 waste packages each) are represented. Two theoretical heat deposit functions have been used representative of waste leading to early and late temperature peaks. This paper presents the thermal code SYRTHES used to simulate large model for transient lasting up to 4000 years. Results show that clay temperature surrounding cells located on the periphery are cooler than those located near the center of the module. For early peaks a reduced domain is well adapted for optimization studies, while for late peak wastes, taking into account an extended domain seems interesting since the large case underlines that optimizing on a reduced domain is clearly conservative. Copyright © 2012 by ASME.

Mimouni S.,Électricité de France | Mechitoua N.,Électricité de France | Foissac A.,Électricité de France | Hassanaly M.,INCKA | Ouraou M.,INCKA
Science and Technology of Nuclear Installations | Year: 2011

The present work is focused on the condensation heat transfer that plays a dominant role in many accident scenarios postulated to occur in the containment of nuclear reactors. The study compares a general multiphase approach implemented in NEPTUNE-CFD with a homogeneous model, of widespread use for engineering studies, implemented in Code-Saturne. The model implemented in NEPTUNE-CFD assumes that liquid droplets form along the wall within nucleation sites. Vapor condensation on droplets makes them grow. Once the droplet diameter reaches a critical value, gravitational forces compensate surface tension force and then droplets slide over the wall and form a liquid film. This approach allows taking into account simultaneously the mechanical drift between the droplet and the gas, the heat and mass transfer on droplets in the core of the flow and the condensation/evaporation phenomena on the walls. As concern the homogeneous approach, the motion of the liquid film due to the gravitational forces is neglected, as well as the volume occupied by the liquid. Both condensation models and compressible procedures are validated and compared to experimental data provided by the TOSQAN ISP47 experiment (IRSN Saclay). Computational results compare favorably with experimental data, particularly for the Helium and steam volume fractions. Copyright © 2011 S. Mimouni et al.

Mimouni S.,Électricité de France | Mechitoua N.,Électricité de France | Ouraou M.,INCKA
Science and Technology of Nuclear Installations | Year: 2011

A large amount of Hydrogen gas is expected to be released within the dry containment of a pressurized water reactor (PWR), shortly after the hypothetical beginning of a severe accident leading to the melting of the core. According to local gas concentrations, the gaseous mixture of hydrogen, air and steam can reach the flammability limit, threatening the containment integrity. In order to prevent mechanical loads resulting from a possible conflagration of the gas mixture, French and German reactor containments are equipped with passive autocatalytic recombiners (PARs) which preventively oxidize hydrogen for concentrations lower than that of the flammability limit. The objective of the paper is to present numerical assessments of the recombiner models implemented in CFD solvers NEPTUNE-CFD and Code-Saturne. Under the EDF/EPRI agreement, CEA has been committed to perform 42 tests of PARs. The experimental program named KALI-H2, consists checking the performance and behaviour of PAR. Unrealistic values for the gas temperature are calculated if the conjugate heat transfer and the wall steam condensation are not taken into account. The combined effects of these models give a good agreement between computational results and experimental data. Copyright © 2011 Stéphane Mimouni et al.

Loading INCKA collaborators
Loading INCKA collaborators