Abstract
The primary goal of this paper is to increase the efficiency of criticality and burnup calculations in the ANSWERS MONK® Monte Carlo code [1]. Two ways of achieving this goal are investigated as part of the H2020 McSAFE Project: creating a unified energy grid for all materials in the model, and reducing the spread in variances of fluxes for depletable materials using a generated optimised importance map. The average tracking speedup factor across all cycles of all burnup calculations ran using the unified energy grid, at base temperature, was found to be 1.96. For criticality calculations at 400K with runtime Doppler broadening, the unified grid approach gave a total speedup factor of 7.32. This demonstrates the potential importance of this method to reduce the calculation time with models with runtime Doppler broadening. The use of the generated optimised importance map has been demonstrated to significantly reduce the variance in the standard deviations on the fluxes in the fuel pins across two different test cases. If a solution is required in which the standard deviation in none of the fuel pins exceeds 5% it was found that the number of scoring stages required was more than halved, highlighting the potential for the outlined methodology to speedup burnup credit calculations.
Highlights
At the heart of Monte Carlo (MC) neutronics codes is the interrogation of nuclear data for the purposes of neutron tracking
Unified Energy Grid Calculations vary the thinning parameter to determine its effect on the results
The speedup is determined by dividing the total run time of a version of the code with the modifications outlined in the methodology against a standard version of MONK®
Summary
At the heart of Monte Carlo (MC) neutronics codes is the interrogation of nuclear data for the purposes of neutron tracking. Many modern MC codes (including the UK Monte Carlo code MONK® [1]) use a detailed nuclide dependent continuous energy representation of cross section data. Particle tracking requires repeated calculation of the mean free path of the particles, which is the inverse of the total macroscopic cross section. As the materials in a burnup calculation comprise many nuclides, which have different energy grids, the lookup process can be slow due to the large numbers of loops required over materials and nuclides. During the McSAFE project a unified energy grid for all materials and nuclides when searching for the required nuclear data was found to invoke a speedup to this process [2]. A suitable algorithm for formulating this unified energy grid is outlined in the methodology
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