Measurement, simulation and uncertainty quantification of the neutron flux at the McMaster Nuclear Reactor. (February 2021)

• Record Type: Journal Article
• Authors: MacConnachie, Elizabeth L.; Novog, David R.
• Is Part Of: Annals of nuclear energy. Volume 151(2021)
• Publication Date: 2021-02
• Subjects: BOC Beginning of Core -- EOC End of Core -- HFIR High Flux Isotope Reactor -- HPGe High Purity Germanium -- LEU Low-enriched Uranium -- MCNP Monte-Carlo N-Particle -- MGRAC Multi-Group Reactor Analysis Code -- MNR McMaster Nuclear Reactor -- MTR Materials Testing Reactor -- OSCAR Overall System for CAlculation of Reactors -- PPF Power Peaking Factor -- TRIGA Training, Research, Isotopes and General Atomics Reactors -- UQ Uncertainty Quantification -- ZED-2 Zero Energy Deuterium Reactor -- R Activation rate [atoms/s] -- A Activity [1/s] -- T Time constant -- N Number of atoms -- ϕ(E) Flux per unit energy [cm−2 s−1 eV−1] -- Φ Energy-integrated flux [cm−2 s−1] -- σn x(E), Energy dependent cross section for reaction (n, x) [cm2] -- σeff Effective cross section [cm2] -- rx g, Neutron induced reaction rate in isotope x in energy group g [1/s] -- P Reactor power -- Ν Average number of neutrons released per fission event -- Efission Energy released per fission event [MeV] -- E0 Most probable energy for a Maxwellian distribution -- Q Energy conversion factor, Q = 1.6022 × 10−13 J/MeV -- keff Effective multiplication factor -- FCd Cadmium correction factor; McMaster Nuclear Reactor -- Neutron activation analysis -- Nuclear data uncertainty -- MCNP -- Uncertainty quantification; Nuclear energy -- Periodicals; Nuclear engineering -- Periodicals; 621.4805
• DOI: 10.1016/j.anucene.2020.107879
• Languages: English
• ISSNs: 0306-4549

Abstract

Abstract: Neutron flux measurements in research reactors can be used for code validation and optimizing in-core activation procedures. Since the fuel adjacent to an irradiation site undergoes burnup, and may be shuffled, local flux measurements may be subject to an additional source of burnup-dependent uncertainty. It is unfeasible to perform these measurements for all core conditions; therefore, reactor physics codes may provide supplemental flux information. This work includes a validation study of the MCNP6 model of the McMaster Nuclear Reactor (MNR). Irradiations were performed over several months, with an emphasis on uncertainty quantification during data processing. No change in the local flux was measured over this period of operation, indicating that burnup effects may be insignificant compared to other sources of uncertainty. These results were validated by five sets of computational data from historical MNR cores. Burnup effects do not need to be accounted for in determining neutron flux uncertainties.