The Experts below are selected from a list of 294 Experts worldwide ranked by ideXlab platform
Guillaume Ricciardi - One of the best experts on this subject based on the ideXlab platform.
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Fluid-structure interaction modelling of a PWR fuel assembly subjected to axial flow
Journal of Fluids and Structures, 2016Co-Authors: Guillaume RicciardiAbstract:Nuclear industry needs tools to design Reactor Cores in case of earthquake. A fluid-structure model simulating the response of the core to a seismic excitation have been developed. Full scale tests considering one fuel assembly are performed to identify coefficients (added mass and damping) that will be used as inputs in the models. Tests showed that the axial water flow induced an added stiffness. In the paper, an expression of the model accounting for the fluid in the fuel assembly with a porous media model and in the bypasses with a leakage flow model is developed. Numerical simulations are compared to experiments and showed good agreement.
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modelling pressurized water Reactor Cores in terms of porous media
Journal of Fluids and Structures, 2009Co-Authors: Guillaume Ricciardi, Sergio Bellizzi, Bruno Collard, Bruno CochelinAbstract:The aim of this study is to develop a tractable model of a nuclear Reactor core taking the complexity of the structure (including its nonlinear behaviour) and fluid flow coupling into account. The mechanical behaviour modelling includes the dynamics of both the fuel assemblies and the fluid. Each rod bundle is modelled in the form of a deformable porous medium; then, the velocity field of the fluid and the displacement field of the structure are defined over the whole domain. The fluid and the structure are first modelled separately, before being linked together. The equations of motion for the structure are obtained using a Lagrangian approach and, to be able to link up the fluid and the structure, the equations of motion for the fluid are obtained using an arbitrary Lagragian Eulerian approach. The finite element method is applied to spatially discretize the equations. Simulations are performed to analyse the effects of the characteristics of the fluid and of the structure. Finally, the model is validated with a test involving two fuel assemblies, showing good agreement with the experimental data.
Minoru Takahashi - One of the best experts on this subject based on the ideXlab platform.
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design study of small lead cooled fast Reactor Cores using sic cladding and structure
Journal of Power and Energy Systems, 2007Co-Authors: Abu Khalid Rivai, Minoru TakahashiAbstract:Neutronics of a Reactor core with SiC cladding and structure was compared with that with steel cladding and structure analytically for small lead-cooled fast Reactors. Uranium nitride fuel was used for this Reactor. U235 enrichment was 11% in inner core and 13% in outer core for relatively flat neutron flux distributions and power density distribution. The core design was optimized using natural uranium blanket and nitride fuel for long life-core with reshuffling interval of 15 years. The analytical result indicated that neutron energy spectrum was slightly softer in the core with the SiC cladding and structure than that with steel cladding and structure. The SiC type Reactor was designed to have criticality at the beginning of cycle (BOC), although the steel type Reactor could not have criticality with the same size and geometry. In other words, the SiC type core can be designed smaller than the steel type core. The peak power densities could remain constant over the Reactor operation. The consumption capability of uranium was quite high, i.e. 10% for 125 MWt Reactor and 18.4% for 375 MWt Reactor at the end of cycle (EOC).
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Design study of PbBi- and NaK-cooled small deep sea fast Reactors
Progress in Nuclear Energy, 2005Co-Authors: Akira Otsubo, Minoru TakahashiAbstract:Abstract The liquid lead-bismuth eutectic (Pb Bi) has good compatibility with water, which is different from sodium. It is expected that the Pb Bi could be used as a coolant of the deep sea fast Reactor (DSFR). Physics analysis of the Pb Bi-cooled small Reactor Cores with and without inner control rods performed using the computer program of a neutronics code system (SRAC95) shows that Pb Bi is suitable for the coolant of small Reactors as well as NaK.
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Design Study of Pb-Bi-Cooled and NaK-Cooled Small Reactors: PBWFR and DSFR
2004Co-Authors: Akira Otsubo, Minoru TakahashiAbstract:The liquid lead-bismuth eutectic (Pb-Bi) has good compatibility with water, which is different from sodium. It is expected that the Pb-Bi could be used as a coolant of the deep sea fast Reactor (DSFR) and the Pb-Bi- cooled direct contact boiling water small fast Reactor (PBWFR). Physics analysis of the Pb-Bi-cooled small Reactor Cores with and without inner control rods was performed using the computer program of General Purpose Neutronics Code System (SRAC95) developed by Japan Atomic Energy Research Institute (JAERI). The coolant of Pb-Bi seems to be good as well as NaK for small Reactors. (authors)
Andrew R. Siegel - One of the best experts on this subject based on the ideXlab platform.
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A Performance Analysis of SIMD Algorithms for Monte Carlo Simulations of Nuclear Reactor Cores
2015 IEEE International Parallel and Distributed Processing Symposium, 2015Co-Authors: David Ozog, Allen D. Malony, Andrew R. SiegelAbstract:A primary characteristic of history-based Monte Carlo neutron transport simulation is the application of MIMD-style parallelism: the path of each neutron particle is largely independent of all other particles, so threads of execution perform independent instructions with respect to other threads. This conflicts with the growing trend of HPC vendors exploiting SIMD hardware, which accomplishes better parallelism and more FLOPS per watt. Event-based neutron transport suits vectorization better than history-based transport, but it is difficult to implement and complicates data management and transfer. However, the Intel Xeon Phi architecture supports the familiar ×86 instruction set and memory model, mitigating difficulties in vector zing neutron transport codes. This paper compares the event-based and history-based approaches for exploiting SIMD in Monte Carlo neutron transport simulations. For both algorithms, we analyze performance using the three different execution models provided by the Xeon Phi (offload, native, and symmetric) within the full-featured OpenVMS framework. A representative micro-benchmark of the performance bottleneck computation shows about 10x performance improvement using the event-based method. In an optimized history-based simulation of a full-physics nuclear Reactor core in OpenVMS, the MIC shows a calculation rate 1.6x higher than a modern 16-core CPU, 2.5x higher when balancing load between the CPU and 1 MIC, and 4x higher when balancing load between the CPU and 2 Macs. As far as we are aware, our calculation rate per node on a high fidelity benchmark (17, 098 particles/second) is higher than any other Monte Carlo neutron transport application. Furthermore, we attain 95% distributed efficiency when using MPI and up to 512 concurrent MIC devices.
A.w. Krass - One of the best experts on this subject based on the ideXlab platform.
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Experimental Criticality Benchmarks for SNAP 10A/2 Reactor Cores
2005Co-Authors: A.w. KrassAbstract:This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type Reactor Cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type Reactor Cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.
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experimental criticality benchmarks for snap 10a 2 Reactor Cores
Transactions of the American Nuclear Society, 2005Co-Authors: A.w. KrassAbstract:This report describes computational benchmark models for nuclear criticality derived from descriptions of the Systems for Nuclear Auxiliary Power (SNAP) Critical Assembly (SCA)-4B experimental criticality program conducted by Atomics International during the early 1960's. The selected experimental configurations consist of fueled SNAP 10A/2-type Reactor Cores subject to varied conditions of water immersion and reflection under experimental control to measure neutron multiplication. SNAP 10A/2-type Reactor Cores are compact volumes fueled and moderated with the hydride of highly enriched uranium-zirconium alloy. Specifications for the materials and geometry needed to describe a given experimental configuration for a model using MCNP5 are provided. The material and geometry specifications are adequate to permit user development of input for alternative nuclear safety codes, such as KENO. A total of 73 distinct experimental configurations are described.
Bruno Cochelin - One of the best experts on this subject based on the ideXlab platform.
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modelling pressurized water Reactor Cores in terms of porous media
Journal of Fluids and Structures, 2009Co-Authors: Guillaume Ricciardi, Sergio Bellizzi, Bruno Collard, Bruno CochelinAbstract:The aim of this study is to develop a tractable model of a nuclear Reactor core taking the complexity of the structure (including its nonlinear behaviour) and fluid flow coupling into account. The mechanical behaviour modelling includes the dynamics of both the fuel assemblies and the fluid. Each rod bundle is modelled in the form of a deformable porous medium; then, the velocity field of the fluid and the displacement field of the structure are defined over the whole domain. The fluid and the structure are first modelled separately, before being linked together. The equations of motion for the structure are obtained using a Lagrangian approach and, to be able to link up the fluid and the structure, the equations of motion for the fluid are obtained using an arbitrary Lagragian Eulerian approach. The finite element method is applied to spatially discretize the equations. Simulations are performed to analyse the effects of the characteristics of the fluid and of the structure. Finally, the model is validated with a test involving two fuel assemblies, showing good agreement with the experimental data.
