A General Method of Conserving Mass in Complex Geometry Simulations on Mesh Grids and Its Implementation in BOT3P5.0

Orsi, Roberto

doi:10.13182/nse06-a2631

Nuclear Science and Engineering

2006

DOI: 10.13182/nse06-a2631

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A General Method of Conserving Mass in Complex Geometry Simulations on Mesh Grids and Its Implementation in BOT3P5.0

Roberto Orsi

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Cited by 4 publications

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“…The construction of grand approximations for the geometry of the problem which support local (in each difference cell) or global (within bodies or a confrontational region) mass balance has already been examined in [7].…”

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confidence: 99%

Calculation of neutron fields in a reactor core using approximations maintaining materials mass balance in a finite-difference cell grid

et al. 2008

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The possibility of using geometrical approximations, maintaining local materials mass balance in each spatial grid cell by introducing additional mixtures for the cells where several initial materials are present, in kinematic calculations of neutron fields in a reactor core is analyzed. To prescribe a 3D geometry of the core, combinatorial geometry methods, implemented in the MCU computer program for obtaining a Monte Carlo solution of the transport equation, are used to convert the combinatorial formulation of the geometry into a grid representation -the ray tracing method. Calculations of a VVER-1000 core and a model of a spent fuel repository show that the method considered here gives a severalfold computational gain over standard approximations of the geometry.A direct kinetic calculation of neutron fields in a reactor is a topical problem, since it permits lifting the well-known limitations of the diffusion approximation and thereby increasing substantially the accuracy of calculations of the energy release, fuel burnup, and other characteristics which are of practical interest. Difficulties arise in such calculations because it is necessary to prescribe a three-dimensional geometry of the computational region, convert the geometry to a 3D difference grid with several million cells, and use efficient methods to accelerate internal iterations over the scattering integral in a group and external iterations over the region of neutron thermalization and the fission source when determining the effective neutron multiplication coefficient k eff . Another condition for obtaining the desired computational accuracy is the availability of a high-accuracy multigroup or problem-oriented system of constants, making it possible to take account of heterogeneity, to the desired degree of accuracy, when calculating the resonance blocking of cross-sections in fuel elements and other structural components of the core.As a result of the high heterogeneity of the core, attempts to use standard approximations of the geometry on a grid, including irregular grids, require for convergence dense three-dimensional grids with spacing 0.1 mm or less, which results in an unjustifiable increase in computational expenditures. The present article analyzes the possibility of using sparser three-dimensional grids by using approximations which maintain local materials mass balance (i.e., in each spatial cell of a grid covering the computational region) by introducing additional mixtures in cells with several initial materials. In this approach, known as the VF (Volume Fraction) method [1, 2], homogenization is achieved at the level of the difference cell of a grid and the degree of homogenization can always be decreased to an acceptable value depending on the required computational accuracy and resources.

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confidence: 99%

Calculation of neutron fields in a reactor core using approximations maintaining materials mass balance in a finite-difference cell grid

et al. 2008

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“…The construction of grid approximations of a geometry which support local (in each spatial grid) or global (within bodies or a computational region) balance of materials mass has also been examined in [8]. The present work examines the effect of local conservation of mass balance on the accuracy of the calculation of radiation fields in the shielding of a VVER-1500 reactor, which is now being designed.…”

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Calculation of radiation fields in VVER shielding using approximations supporting local balance of materials mass and fission-source neutrons

et al. 2008

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The possibility of using approximations of geometry, which support the balance of materials mass on each spatial cell of a grid covering the computational region, to calculate radiation fields in VVER shielding by introducing additional mixtures of materials for cells where several initial materials are present is analyzed. The methods of combinatorial geometry, which are implemented in the MCU computer program for solving transfer equations by the Monte Carlo method, are used to specify the 3-D geometry of the shielding and a fuel-element and (or) fuel assembly fission source and the method of ray tracing is used to convert the combinatorial specification into a grid specification. A numerical experiment in which the VVER-1500 radiation shielding is calculated showed that the conservation of the materials mass/number of source neutrons in the grid cells during the conversion ensures rapid convergence of the results when the density of the spatial grid increases. Good agreement is attained between MCU Monte Carlo and KATRIN S n calculations of VVER-1000 shielding.Determining with the required accuracy the radiation fields in VVER shielding to substantiate anticipated regimes and the service life of the vessel and in-vessel apparatus as well as the standard operating regime of sensors for thermal neutrons in the channels of ionization chambers is a topical problem. This problem can be solved by performing a three-dimensional calculation of the real composition by the S n method using, for example, the VITAMIN-B6 multigoup cross-sections library obtained from the BGL440 and BGL1000 problem-oriented libraries [1] of group cross-sections with 47 groups of neutrons and 20 groups of photons, taking account of neutron thermalization, as well as data on fuel burnup (which are given for a fuel elements and/or fuel assemblies), whose profile repeats the time-integrated profile of the fission-neutron source. The difficulties encountered in such calculations are due to the necessity of prescribing the three-dimensional geometry of the computational region and converting the geometry and the source to a difference grid using spatial grids with several million cells and effective methods for accelerating the inner and outer (with respect to the region of neutron thermalization) iterations.

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Potential Enhanced Performances in Radiation Transport Analysis on Structured Mesh Grids Made Available by BOT3P

Orsi

2007

Nuclear Science and Engineering

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A General Method of Conserving Mass in Complex Geometry Simulations on Mesh Grids and Its Implementation in BOT3P5.0

Cited by 4 publications

References 5 publications

Calculation of neutron fields in a reactor core using approximations maintaining materials mass balance in a finite-difference cell grid

Calculation of neutron fields in a reactor core using approximations maintaining materials mass balance in a finite-difference cell grid

Calculation of radiation fields in VVER shielding using approximations supporting local balance of materials mass and fission-source neutrons

Potential Enhanced Performances in Radiation Transport Analysis on Structured Mesh Grids Made Available by BOT3P

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