Abstract
Safe operation of a nuclear reactor requires an extensive calculational support. Operational data are determined by full-core calculations during the design phase of a fuel loading. Loading pattern and design of fuel assemblies are adjusted to meet safety requirements and optimize reactor operation. Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors. Nuclear data for this diffusion code are prepared regularly by lattice code HELIOS. These calculations are conducted in 2D on fuel assembly level. There is also possibility to calculate these macroscopic data by Monte-Carlo Serpent code. It can make use of alternative evaluated libraries. All calculations are affected by inherent uncertainties in nuclear data. It is useful to see results of full-core calculations based on two sets of diffusion data obtained by Serpent code calculations with ENDF/B-VII.1 and JEFF-3.2 nuclear data including also decay data library and fission yields data. The comparison is based directly on fuel assembly level macroscopic data and resulting operational data. This study illustrates effect of evaluated nuclear data library on full-core calculations of a large PWR reactor core. The level of difference which results exclusively from nuclear data selection can help to understand the level of inherent uncertainties of such full-core calculations.
Highlights
Nodal diffusion code ANDREA is used for this task in case of Czech VVER-1000 reactors
The typical sequence of neutronic calculations of nuclear reactors starts with preparation of macroscopic data by detailed lattice calculations on the fuel assembly level
It can be explained by about 25% lower yield of 156Gd by thermal fission of 235U for JEFF-3.2 compared to ENDF/B-VII.1, which results in lower production of 157Gd by subsequent neutron radiative capture
Summary
The typical sequence of neutronic calculations of nuclear reactors starts with preparation of macroscopic data by detailed lattice calculations on the fuel assembly level. These data are later applied in full-core calculations by deterministic methods. The last one employes Monte-Carlo method and allows easy switching of input nuclear data It uses ACE format [2] for continuous cross-section data and directly evaluated data for decay and fission yields data. It allows demonstration of isolated impact of nuclear data on macroscopic data and full-core calculations
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