Evaluation and enhancement of COBRA-TF efficiency for LWR calculations

Cuervo, D.; Avramova, Maria; Ivanov, Kostadin; Miró, R.

doi:10.1016/j.anucene.2006.03.011

Cited by 15 publications

(9 citation statements)

References 7 publications

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“…3.4.1 COBRA-TF COBRA-TF (CTF) includes a rod-bundle subchannel thermal-hydraulics application used to perform transient simulation for the full range of two-phase flow regimes based upon the geometry inherent in current commercial nuclear power reactors [44]. The subchannel approach can be thought of as an extremely coarse-mesh approximation, in which the control volume is of a size equivalent to a single fuel rod and its surrounding coolant.…”

Section: Physics Components Selected For Veramentioning

confidence: 99%

The Virtual Environment for Reactor Applications (VERA): Design and architecture

Turner

Clarno

Sieger

et al. 2016

Journal of Computational Physics

122

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VERA, the Virtual Environment for Reactor Applications, is the system of physics capabilities being developed and deployed by the Consortium for Advanced Simulation of Light Water Reactors (CASL). CASL was established for the modeling and simulation of commercial nuclear reactors. VERA consists of integrating and interfacing software together with a suite of physics components adapted and/or refactored to simulate relevant physical phenomena in a coupled manner. VERA also includes the software development environment and computational infrastructure needed for these components to be effectively used. We describe the architecture of VERA from both software and numerical perspectives, along with the goals and constraints that drove major design decisions, and their implications. We explain why VERA is an environment rather than a framework or toolkit, why these distinctions are relevant (particularly for coupled physics applications), and provide an overview of results that demonstrate the use of VERA tools for a variety of challenging applications within the nuclear industry.

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Section: Physics Components Selected For Veramentioning

confidence: 99%

The Virtual Environment for Reactor Applications (VERA): Design and architecture

Turner

Clarno

Sieger

et al. 2016

Journal of Computational Physics

122

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“…In this section, we provide a brief overview of this code, and interested readers can find further details in [5].…”

Section: Description Of Ctfmentioning

confidence: 99%

Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

Ramos

Román

Abarca

et al. 2016

Annals of Nuclear Energy

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In order to reduce the response time when simulating large reactors in detail, a parallel version of the thermal-hydraulic subchannel code COBRA-TF (CTF) has been developed using the standard Message Passing Interface (MPI). The parallelization is oriented to reactor cells, so it is best suited for models consisting of many cells. The generation of the Jacobian matrix is parallelized, in such a way that each processor is in charge of generating the data associated with a subset of cells. Also, the solution of the linear system of equations is done in parallel, using the PETSc toolkit. With the goal of $ This work has been partially supported by the Universitat Politècnica de València under Projects COBRA PAR (PAID-05-11-2810) and OpenNUC (PAID-05-12), and by the Spanish Ministerio de Economa y Competitividad under Projects SLEPc-PFE (TIN2013-41049-P) and NUC-MULTPHYS (ENE2012-34585). The authors would like to acknowledge the technical support provided by CNAT and IBERDROLA GENERACIÓN S.A. for the realization of this work.* Corresponding author. creating a powerful tool to simulate the reactor core behavior during asymmetrical transients, the 3D neutron diffusion code PARCSv2.7 (PARCS) has been coupled with the parallel version of CTF (pCTF) using the Parallel Virtual Machine (PVM) technology. In order to validate the correctness of the parallel coupled code, a control rod drop transient has been simulated comparing the results with the real experimental measures acquired during an NPP real test.

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“…In the steady-state 1/8 symmetry PWR core geometry (26 assemblies Â 50 nodes = 1,300 meshes), the calculation time of COBRA-TF is 220 s using a single CPU (CPU: Pentium IV 3.2 GHz, Memory: 512 MB). 35) In the BWR fuel assembly that is used in this study, the number of meshes in each assembly is approximately 1,900 (9 Â 9 subchannels Â 24 nodes).…”

Section: Estimation Of Calculation Time Of Subaru Thmentioning

confidence: 99%

Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations

Tada

Fujita

Endo

et al. 2011

Journal of Nuclear Science and Technology

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Three-dimensional pin-by-pin core analysis is considered to be a candidate for the next-generation BWR core calculation method. In our previous study, the applicability of the transport and burnup calculations for a three-dimensional pin-by-pin BWR core analysis was investigated. However, the thermal-hydraulics calculation has not yet been studied in this framework. In the conventional core analysis code, the bundlewise thermal-hydraulics calculation is adopted. In the actual core analysis, the power distribution inside a fuel assembly is tilted at the region adjacent to a control blade or the core peripheral region. In these regions, the consideration of the subchannel-wise void distribution has an impact on the fission rate distribution. Therefore, an evaluation of the detailed void distribution inside an assembly, i.e., the incorporation of the subchannel wise void distribution, is desirable for the pin-by-pin BWR core analysis. Although several subchannel analysis codes have been developed, these subchannel analysis codes generally require a large computational effort to estimate the subchannel-wise void distribution in a whole BWR core. Therefore, to analyze a whole BWR core within a reasonable computation time, it was necessary to apply a fast subchannel analysis code. In this paper, a quick subchannel analysis code dedicated to pin-by-pin BWR core analysis is newly developed, and the void distribution of the present subchannel analysis code is compared with the prevailing subchannel analysis code NASCA using three-dimensional single-assembly geometries. Since the present subchannel analysis code is used for a coupled neutronics/thermal-hydraulics analysis, the results of the coupling calculation are also compared with those of NASCA. The calculation result indicates that the void distribution difference between NASCA and the present subchannel analysis code is slightly less than 10%. This result indicates that the prediction accuracy of the present subchannel analysis code will be reasonably appropriate for a pin-by-pin BWR core analysis. Furthermore, the results show that the calculation time of the present subchannel analysis code is only 10 min for a hypothetical three-dimensional ABWR quarter-core geometry using a single CPU. This calculation time is sufficient for a pin-by-pin BWR core analysis.

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Evaluation and enhancement of COBRA-TF efficiency for LWR calculations

Cited by 15 publications

References 7 publications

The Virtual Environment for Reactor Applications (VERA): Design and architecture

The Virtual Environment for Reactor Applications (VERA): Design and architecture

Control rod drop transient analysis with the coupled parallel code pCTF-PARCSv2.7

Application of Quick Subchannel Analysis Method for Three-Dimensional Pin-by-Pin BWR Core Calculations

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