| 1. |
- Dufek, Jan, 1978-, et al.
(author)
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Monte Carlo criticality calculations accelerated by a growing neutron population
- 2016
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In: Annals of Nuclear Energy. - : Elsevier. - 0306-4549 .- 1873-2100. ; 94, s. 16-21
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Journal article (peer-reviewed)abstract
- We propose a fission source convergence acceleration method for Monte Carlo criticality simulation. As the efficiency of Monte Carlo criticality simulations is sensitive to the selected neutron population size, the method attempts to achieve the acceleration via on-the-fly control of the neutron population size. The neutron population size is gradually increased over successive criticality cycles so that the fission source bias amounts to a specific fraction of the total error in the cumulative fission source. An optimal setting then gives a reasonably small neutron population size, allowing for an efficient source iteration; at the same time the neutron population size is chosen large enough to ensure a sufficiently small source bias, such that does not limit accuracy of the simulation.
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| 2. |
- Mickus, Ignas, et al.
(author)
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PERFORMANCE OF THE EXPLICIT EULER AND PREDICTOR-CORRECTOR-BASED COUPLING SCHEMES IN MONTE CARLO BURNUP CALCULATIONS OF FAST REACTORS
- 2015
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In: Nuclear Technology. - 0029-5450 .- 1943-7471. ; 191:2, s. 193-198
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Journal article (peer-reviewed)abstract
- We present a stability test of the explicit Euler and predictor-corrector based coupling schemes in Monte Carlo burnup calculations of the gas fast reactor fuel assembly. Previous studies have identified numerical instabilities of these coupling schemes in Monte Carlo burnup calculations of thermal-spectrum reactors due to spatial feedback induced neutron flux and nuclide density oscillations, where only sufficiently small time steps could guarantee acceptable precision. New results suggest that these instabilities are insignificant in fast-spectrum assembly burnup calculations, and the considered coupling schemes can therefore perform well in fast-spectrum reactor burnup calculations even with relatively large time steps. Note: Some figures in this technical note may be in color only in the electronic version.
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| 3. |
- Tuttelberg, Kaur, et al.
(author)
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Estimation of errors in the cumulative Monte Carlo fission source
- 2014
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In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 72, s. 151-155
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Journal article (peer-reviewed)abstract
- We study the feasibility of estimating the error in the cumulative fission source in Monte Carlo criticality calculations by utilising the fundamental-mode eigenvector of the fission matrix. The cumulative fission source, representing the source combined over active cycles, contains errors of both statistical and systematic nature. Knowledge of the error in the cumulative fission source is crucial as it determines the accuracy of computed neutron flux and power distributions. While statistical errors are present in the eigenvector of the fission matrix, it appears that these are not (or they are only weakly) correlated to the errors in the cumulative fission source. This ensures the suggested methodology gives error estimates that are distributed around the real errors, which is also supported by results of our numerical test calculations.
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| 4. |
- Tuttelberg, Kaur, et al.
(author)
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Neutron batch size optimisation methodology for Monte Carlo criticality calculations
- 2015
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In: Annals of Nuclear Energy. - : Elsevier BV. - 0306-4549 .- 1873-2100. ; 75, s. 620-626
-
Journal article (peer-reviewed)abstract
- We present a methodology that improves the efficiency of conventional power iteration based Monte Carlo criticality calculations by optimising the number of neutron histories simulated per criticality cycle (the so-called neutron batch size). The chosen neutron batch size affects both the rate of convergence (in computing time) and magnitude of bias in the fission source. Setting a small neutron batch size ensures a rapid simulation of criticality cycles, allowing the fission source to converge fast to its stationary state; however, at the same time, the small neutron batch size introduces a large systematic bias in the fission source. It follows that for a given allocated computing time, there is an optimal neutron batch size that balances these two effects. We approach this problem by studying the error in the cumulative fission source, i.e. the fission source combined over all simulated cycles, as all results are commonly combined over the simulated cycles. We have deduced a simplified formula for the error in the cumulative fission source, taking into account the neutron batch size, the dominance ratio of the system, the error in the initial fission source and the allocated computing time (in the form of the total number of simulated neutron histories). Knowing how the neutron batch size affects the error in the cumulative fission source allows us to find its optimal value. We demonstrate the benefits of the method on a number of numerical test calculations.
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