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Research on Coupling Scheme of Monte Carlo Burnup Calculation in RMC

[+] Author Affiliations
Wanlin Li, Kan Wang, Ganglin Yu, Yaodong Li

Tsinghua University, Beijing, China

Paper No. ICONE26-81140, pp. V003T02A005; 6 pages
doi:10.1115/ICONE26-81140
From:
  • 2018 26th International Conference on Nuclear Engineering
  • Volume 3: Nuclear Fuel and Material, Reactor Physics, and Transport Theory
  • London, England, July 22–26, 2018
  • Conference Sponsors: Nuclear Engineering Division
  • ISBN: 978-0-7918-5145-6
  • Copyright © 2018 by ASME

abstract

Monte Carlo (MC) burnup calculation method, implemented through coupling neutron transport and point depletion solvers, is widely used in design and analysis of nuclear reactor. Burnup calculation is generally solved by dividing reactor lifetime into steps and modeling geometry into numbers of burnup areas where neutron flux and one group effective cross sections are treated as constant during each burnup step. Such constant approximation for neutron flux and effective cross section will lead to obvious error unless using fairly short step. To yield accuracy and efficiency improvement, coupling schemes have been researched in series of MC codes.

In this study, four coupling schemes, beginning of step approximation, predictor-corrector methods by correcting nuclide density and flux-cross section as well as high order predictor-corrector with sub-step method were researched and implemented in RMC. Verification and comparison were performed by adopting assembly problem from VERA international benchmark. Results illustrate that high order coupled with sub-step method is with notable accuracy compared to beginning of step approximation and traditional predictor-corrector, especially for calculation in which step length is fairly long.

Copyright © 2018 by ASME

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