Institute Of Genie Nucleaire

Montréal, Canada

Institute Of Genie Nucleaire

Montréal, Canada
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Chambon R.,Institute Of Genie Nucleaire | Hebert A.,Institute Of Genie Nucleaire | Taforeau J.,Institute for Radiological Protection and Nuclear Safety
International Journal of Nuclear Energy Science and Technology | Year: 2017

In order to better optimise the fuel energy efficiency and perform safety analyses in PWRs, the fuel power distribution has to be known as accurately as possible, ideally in each pin. However, this level of detail is lost when core calculations are performed with cross-sections homogenised over the fuel assembly. The pin power reconstruction (PPR) method can be used to get back this level of detail as accurately as possible in a small additional computing time frame compared to pin-by-pin full-core calculations. The DRAGON5 lattice code and the PARCS core code were recently interfaced. For this study, all the missing parts to be able to perform PPR were introduced in the newly developed system DRAGON5/PARCS. A major component was to set the methodology to compute the corner and assembly discontinuity factors in DRAGON5. Verification tests were performed on 12 configurations of 3x3 clusters where simulations in transport theory and in diffusion theory followed by pin-power reconstruction were compared. © 2017 Inderscience Enterprises Ltd.


Bernal A.,Polytechnic University of Valencia | Hebert A.,Institute Of Genie Nucleaire | Roman J.E.,Polytechnic University of Valencia | Miro R.,Polytechnic University of Valencia | Verdu G.,Polytechnic University of Valencia
Journal of Nuclear Science and Technology | Year: 2017

Mixed-dual formulations of the finite element method were successfully applied to the neutron diffusion equation, such as the Raviart–Thomas method in Cartesian geometry and the Raviart–Thomas–Schneider in hexagonal geometry. Both methods obtain system matrices which are suitable for solving the eigenvalue problem with the preconditioned power method. This method is very fast and optimized, but only for the calculation of the fundamental mode. However, the determination of non-fundamental modes is important for modal analysis, instabilities, and fluctuations of nuclear reactors. So, effective and fast methods are required for solving eigenvalue problems. The most effective methods are those based on Krylov subspaces projection combined with restart, such as Krylov–Schur. In this work, a Krylov–Schur method has been applied to the neutron diffusion equation, discretized with the Raviart–Thomas and Raviart–Thomas–Schneider methods. © 2017 Atomic Energy Society of Japan. All rights reserved.


Clerc T.,Institute Of Genie Nucleaire | Hebert A.,Institute Of Genie Nucleaire | Leroyer H.,Électricité de France | Argaud J.P.,Électricité de France | And 2 more authors.
Nuclear Engineering and Design | Year: 2014

This paper presents a computational scheme for the determination of equivalent 2D multi-group spatially dependant reflector parameters in a Pressurized Water Reactor (PWR). The proposed strategy is to define a full-core calculation consistent with a reference lattice code calculation such as the Method Of Characteristics (MOC) as implemented in APOLLO2 lattice code. The computational scheme presented here relies on the data assimilation module known as "Assimilation de données et Aide à l'Optimisation (ADAO)" of the SALOME platform developed at Électricité De France (EDF), coupled with the full-core code COCAGNE and with the lattice code APOLLO2. A first code-to-code verification of the computational scheme is made using the OPTEX reflector model developed at École Polytechnique de Montréal (EPM). As a result, we obtain 2D multi-group, spatially dependant reflector parameters, using both diffusion or SPN operators. We observe important improvements of the power discrepancies distribution over the core when using reflector parameters computed with the proposed computational scheme, and the SPN operator enables additional improvements. © 2014 Elsevier B.V.


Chambon R.,Institute Of Genie Nucleaire | Hebert A.,Institute Of Genie Nucleaire
Nuclear Engineering and Design | Year: 2015

In order to better optimize the fuel energy efficiency in PWRs, the burnup distribution has to be known as accurately as possible, ideally in each pin. However, this level of detail is lost when core calculations are performed with homogenized cross-sections. The pin power reconstruction (PPR) method can be used to get back those levels of details as accurately as possible in a small additional computing time frame compared to classical core calculations. Such a de-homogenization technique for core calculations using arbitrarily homogenized fuel assembly geometries was presented originally by Fliscounakis et al. In our work, the same methodology was implemented in the open-source neutronic codes DRAGON5 and DONJON5. The new type of Selengut homogenization, called macro-calculation water gap, also proposed by Fliscounakis et al. was implemented. Some important details on the methodology were emphasized in order to get precise results. Validation tests were performed on 12 configurations of 3×3 clusters where simulations in transport theory and in diffusion theory followed by pin-power reconstruction were compared. The results shows that the pin power reconstruction and the Selengut macro-calculation water gap methods were correctly implemented. The accuracy of the simulations depends on the SPH method and on the homogenization geometry choices. Results show that the heterogeneous homogenization is highly recommended. SPH techniques were investigated with flux-volume and Selengut normalization, but the former leads to inaccurate results. Even though the new Selengut macro-calculation water gap method gives promising results regarding flux continuity at assembly interfaces, the classical Selengut approach is more reliable in terms of maximum and average errors in the whole range of configurations. © 2015 Elsevier B.V. All rights reserved.


Clerc T.,Institute Of Genie Nucleaire | Hebert A.,Institute Of Genie Nucleaire | Leroyer H.,Électricité de France | Argaud J.-P.,Électricité de France | And 2 more authors.
International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2013 | Year: 2013

This paper presents a work-flow for computing an equivalent 2D radial reflector in a pressurized water reactor (PWR) core, in adequacy with a reference power distribution, computed with the method of caracteristics (MOC) of the lattice code APOLLO2. The Multi-modelling Equivalent Reflector Computation (MERC) work-flow is a coherent association of the lattice code APOLLO2 and the core code COCAGNE, structured around the ADAO (Assimilation de Données et Aide à l'Optimisation) module of the SALOME platform, based on the data assimilation theory. This study leads to the computation of equivalent few-groups reflectors, that can be spatially heterogeneous, which have been compared to those obtained with the OPTEX similar methodology developed with the core code DONJON, as a first validation step. Subsequently, the MERC work-flow is used to compute the most accurate reflector in consistency with all the R&D choices made at Électricité de France (EDF) for the core modelling, in terms of number of energy groups and simplified transport solvers. We observe important reductions of the power discrepancies distribution over the core when using equivalent reflectors obtained with the MERC work-flow.

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