Optimal time step length and statistics in Monte Carlo burnup simulations

Dufek, Jan; Mickus, Ignas

doi:10.1016/j.anucene.2019.107244

Cited by 3 publications

(3 citation statements)

References 12 publications

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“…2, and performed a large number of test simulations with varied values of the free parameters. The test results made it possible for us to suggest an optimisation algorithm of the free parameters [9]. In this paper, we show results not reported previously (see Sec.…”

Section: Introductionsupporting

confidence: 61%

“…All simulations with x close to 0 or 1 underperformed. This work is an extension of our previous study [9,8]. Test results suggest that efficiency of Monte Carlo burnup simulations may be maximised when the fraction of the cost of all depletion calculations (and all other procedures not directly involved in the simulation of neutron histories) in the overall cost of the whole Monte Carlo burnup simulation, x, is inside a certain interval that we approximately evaluated (using a simple numerical model) to be (0.2, 0.8).…”

Section: Test #2: Optimal Conditions For Monte Carlo Burnup Simulationsmentioning

confidence: 61%

“…We have previously studied this optimisation problem. In [9], we have set up a very simplified solver, also briefly described here, in Sec. 2, and performed a large number of test simulations with varied values of the free parameters.…”

Section: Introductionmentioning

confidence: 99%

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Optimisation of Monte Carlo Burnup Simulations

Dufek

Mickus

2021

EPJ Web Conf.

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We show here that computing efficiency of Monte Carlo burnup simulations depends on chosen values of certain free parameters, such as the length of the time steps and the number of neutron histories simulated at each Monte Carlo criticality run. The efficiency can thus be improved by optimising these parameters. We have set up a simple numerical model that made it possible for us to test a large number of combinations of the free parameters, and suggest a way to optimise their selection.

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Section: Introductionsupporting

confidence: 61%

Section: Test #2: Optimal Conditions For Monte Carlo Burnup Simulationsmentioning

confidence: 61%

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Optimisation of Monte Carlo Burnup Simulations

Dufek

Mickus

2021

EPJ Web Conf.

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THE McSAFE PROJECT - HIGH-PERFORMANCE MONTE CARLO BASED METHODS FOR SAFETY DEMONSTRATION: FROM PROOF OF CONCEPT TO INDUSTRY APPLICATIONS

et al. 2021

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The increasing use of Monte Carlo methods for core analysis is fostered by the huge and cheap computer power available nowadays e.g. in large HPC. Apart from the classical criticality calculations, the application of Monte Carlo methods for depletion analysis and cross section generation for diffusion and transport core simulators is also expanding. In addition, the development of multi-physics codes by coupling Monte Carlo solvers with thermal hydraulic codes (CFD, subchannel and system thermal hydraulics) to perform full core static core analysis at fuel assembly or pin level has progressed in the last decades. Finally, the extensions of the Monte Carlo codes to describe the behavior of prompt and delay neutrons, control rod movements, etc. has been started some years ago. Recent coupling of dynamic versions of Monte Carlo codes with subchannel codes make possible the analysis of transient e.g. rod ejection accidents and it paves the way for the simulation of any kind of design basis accidents as an alternative option to the use of diffusion and transport based deterministic solvers. The H2020 McSAFE Project is focused on the improvement of methods for depletion considering thermal hydraulic feedbacks, extension of the coupled neutronic/thermal hydraulic codes by the incorporation of a fuel performance solver, the development of dynamic Monte Carlo codes and the development of methods to handle large depletion problems and to reduce the statistical uncertainty. The validation of the multi-physics tools developed within McSAFE will be performed using plant data and unique tests e.g. the SPERT III E REA test. This paper will describe the main developments, solution approaches, and selected results.

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An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

Rochman,

Algora,

Àlvarez-Velarde

et al. 2024

EPJ Nuclear Sci. Technol.

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This paper summarized the efforts performed to understand decay heat estimation from existing spent nuclear fuel (SNF), under the auspices of the Working Party on Nuclear Criticality Safety (WPNCS) of the OECD Nuclear Energy Agency. Needs for precise estimations are related to safety, cost, and optimization of SNF handling, storage, and repository. The physical origins of decay heat (a more correct denomination would be decay power) are then introduced, to identify its main contributors (fission products and actinides) and time-dependent evolution. Due to limited absolute prediction capabilities, experimental information is crucial; measurement facilities and methods are then presented, highlighting both their relevance and our need for maintaining the unique current full-scale facility and developing new ones. The third part of this report is dedicated to the computational aspect of the decay heat estimation: calculation methods, codes, and validation. Different approaches and implementations currently exist for these three aspects, directly impacting our capabilities to predict decay heat and to inform decision-makers. Finally, recommendations from the expert community are proposed, potentially guiding future experimental and computational developments. One of the most important outcomes of this work is the consensus among participants on the need to reduce biases and uncertainties for the estimated SNF decay heat. If it is agreed that uncertainties (being one standard deviation) are on average small (less than a few percent), they still substantially impact various applications when one needs to consider up to three standard deviations, thus covering more than 95% of cases. The second main finding is the need of new decay heat measurements and validation for cases corresponding to more modern fuel characteristics: higher initial enrichment, higher average burnup, as well as shorter and longer cooling time. Similar needs exist for fuel types without public experimental data, such as MOX, VVER, or CANDU fuels. A third outcome is related to SNF assemblies for which no direct validation can be performed, representing the vast majority of cases (due to the large number of SNF assemblies currently stored, or too short or too long cooling periods of interest). A few solutions are possible, depending on the application. For the final repository, systematic measurements of quantities related to decay heat can be performed, such as neutron or gamma emission. This would provide indications of the SNF decay heat at the time of encapsulation. For other applications (short- or long-term cooling), the community would benefit from applying consistent and accepted recommendations on calculation methods, for both decay heat and uncertainties. This would improve the understanding of the results and make comparisons easier.

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Optimal time step length and statistics in Monte Carlo burnup simulations

Cited by 3 publications

References 12 publications

Optimisation of Monte Carlo Burnup Simulations

Optimisation of Monte Carlo Burnup Simulations

THE McSAFE PROJECT - HIGH-PERFORMANCE MONTE CARLO BASED METHODS FOR SAFETY DEMONSTRATION: FROM PROOF OF CONCEPT TO INDUSTRY APPLICATIONS

An introduction to Spent Nuclear Fuel decay heat for Light Water Reactors: a review from the NEA WPNCS

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