Depletion capabilities in the OpenMC Monte Carlo particle transport code
Abstract
A depletion solver has been implemented in OpenMC and is described herein. The depletion solver is implemented in Python and interfaces with OpenMC’s transport solver through a C++ application programming interface, which enables an in-memory transport-depletion coupling. Multiple integration methods for advancing in time have been implemented and exhibit tradeoffs in cost, accuracy, and memory use. For all time integration methods, evaluation of the matrix exponential is performed by using the incomplete partial fraction form of the Chebyshev rational approximation method. Simulations of a pressurized water reactor (PWR) pincell and a sodium-cooled fast reactor (SFR) assembly were carried out with OpenMC and Serpent. For both problems, the use of a high-fidelity depletion chain results in predictions of keff that agree within 20–30 pcm between OpenMC and Serpent. Predicted actinide concentrations were found to agree to a fraction of a percent, and most fission product concentrations were found to agree within 1%. Here, the few cases where larger differences were observed can be attributed either to differences in how the energy dependence of fission product yields is handled or deficiencies in the nuclear data used. OpenMC simulations of the PWR and SFR problems using a simplified 228-nuclide depletion chain demonstrate thatmore »
- Authors:
-
[2];
- Argonne National Lab. (ANL), Argonne, IL (United States)
- Los Alamos National Lab. (LANL), Los Alamos, NM (United States)
- Georgia Inst. of Technology, Atlanta, GA (United States)
- Tsinghua Univ., Beijing (China)
- Publication Date:
- Research Org.:
- Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Argonne National Lab. (ANL), Argonne, IL (United States)
- Sponsoring Org.:
- USDOE National Nuclear Security Administration (NNSA); USDOE Office of Science (SC)
- OSTI Identifier:
- 1726218
- Alternate Identifier(s):
- OSTI ID: 1776818
- Report Number(s):
- LA-UR-20-25299
Journal ID: ISSN 0306-4549; TRN: US2205088
- Grant/Contract Number:
- 89233218CNA000001; 17-SC-20-SC; AC02-06CH11357; AC07-05ID14517
- Resource Type:
- Accepted Manuscript
- Journal Name:
- Annals of Nuclear Energy (Oxford)
- Additional Journal Information:
- Journal Name: Annals of Nuclear Energy (Oxford); Journal Volume: 152; Journal Issue: C; Journal ID: ISSN 0306-4549
- Publisher:
- Elsevier
- Country of Publication:
- United States
- Language:
- English
- Subject:
- 73 NUCLEAR PHYSICS AND RADIATION PHYSICS; Monte Carlo; particle transport; depletion; transmutation
Citation Formats
Romano, Paul K., Josey, Colin J., Johnson, Andrew E., and Liang, Jingang. Depletion capabilities in the OpenMC Monte Carlo particle transport code. United States: N. p., 2020.
Web. doi:10.1016/j.anucene.2020.107989.
Romano, Paul K., Josey, Colin J., Johnson, Andrew E., & Liang, Jingang. Depletion capabilities in the OpenMC Monte Carlo particle transport code. United States. https://doi.org/10.1016/j.anucene.2020.107989
Romano, Paul K., Josey, Colin J., Johnson, Andrew E., and Liang, Jingang. Sat .
"Depletion capabilities in the OpenMC Monte Carlo particle transport code". United States. https://doi.org/10.1016/j.anucene.2020.107989. https://www.osti.gov/servlets/purl/1726218.
@article{osti_1726218,
title = {Depletion capabilities in the OpenMC Monte Carlo particle transport code},
author = {Romano, Paul K. and Josey, Colin J. and Johnson, Andrew E. and Liang, Jingang},
abstractNote = {A depletion solver has been implemented in OpenMC and is described herein. The depletion solver is implemented in Python and interfaces with OpenMC’s transport solver through a C++ application programming interface, which enables an in-memory transport-depletion coupling. Multiple integration methods for advancing in time have been implemented and exhibit tradeoffs in cost, accuracy, and memory use. For all time integration methods, evaluation of the matrix exponential is performed by using the incomplete partial fraction form of the Chebyshev rational approximation method. Simulations of a pressurized water reactor (PWR) pincell and a sodium-cooled fast reactor (SFR) assembly were carried out with OpenMC and Serpent. For both problems, the use of a high-fidelity depletion chain results in predictions of keff that agree within 20–30 pcm between OpenMC and Serpent. Predicted actinide concentrations were found to agree to a fraction of a percent, and most fission product concentrations were found to agree within 1%. Here, the few cases where larger differences were observed can be attributed either to differences in how the energy dependence of fission product yields is handled or deficiencies in the nuclear data used. OpenMC simulations of the PWR and SFR problems using a simplified 228-nuclide depletion chain demonstrate that it achieves accuracy close to that of the full, high-fidelity depletion chain with respect to the studied responses.},
doi = {10.1016/j.anucene.2020.107989},
journal = {Annals of Nuclear Energy (Oxford)},
number = C,
volume = 152,
place = {United States},
year = {Sat Nov 14 00:00:00 EST 2020},
month = {Sat Nov 14 00:00:00 EST 2020}
}
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