QUANTIFICATION OF NUCLEAR DATA AND MANUFACTURING UNCERTAINTIES IN VERA

Christopher Sedota; Scott Palmtag

doi:10.1051/epjconf/202124715014

All issues

Volume 247 (2021)

EPJ Web Conf., 247 (2021) 15014

References

Open Access

Issue		EPJ Web Conf. Volume 247, 2021 PHYSOR2020 – International Conference on Physics of Reactors: Transition to a Scalable Nuclear Future


Article Number		15014
Number of page(s)		8
Section		Sensitivity & Uncertainty Methods
DOI		https://doi.org/10.1051/epjconf/202124715014
Published online		22 February 2021

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[1] B. Kochunas et al., “VERA Core Simulator Methodology for Pressurized Water Reactor Cycle Depletion,” Nucl. Sci. Eng., 185, (2017). [Google Scholar]

[2] B.T. Rearden, M.A. Jessee, “SCALE Code System,” Oak Ridge National Laboratory (2016). [Google Scholar]

[3] K. Ivanov, M. Avramova, S. Kamerow, I. Kodeli, E. Sartori, E. Ivanov, O. Cabellos, “Benchmark for Uncertainty Analysis in Modelling (UAM) for the Design, Operation, and Safety Analysis of LWRs,” NEA/NSC/DOC(2013)7, Nuclear Energy Agency (2013). [Google Scholar]

[4] Ryan N. Bratton, M. Avramova, K. Ivanov, “OECD/NEA Benchmark for Uncertainty Analysis in Modeling (UAM) for LWRs – Summary and Discussion of Neutronics Cases (Phase I),” Nucl. Eng. Tech., 46, Issue 3, pp 313-342 (2014). [Google Scholar]

[5] A. Godfrey, “VERA Core Physics Benchmark Progression Problem Specifications,” Technical Report CASL-U-2012-0131-004, Revision 4 (2014). [Google Scholar]