Conservation of migration area by transport cross sections using Cumulative Migration Method in deterministic heterogeneous reactor transport analysis
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This presentation states that new TSLs in ENDF/B-VIII.0 will reduce the overprediction of the library for the most moderated cases. The differences between VII.1 and VIII.0 are driven by the 239 Pu changes and are consistently higher for all configurations. The VIII.0 239 Pu thermal cross sections were specifically adjusted lower to better predict plutonium thermal solutions criticality benchmarks, yet they overpredict the TEX thermal configurations, moderated by PE. Other, smaller differences have been seen for 240 Pu and 56 Fe. PMM-002 has shown utility in testing cross sections over a wide range of neutron energies.
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Results of neutron dosimetry measurements in ATR for cycle 175C are presented. A single set of cobalt and nickel wires were installed in numerous irradiation locations for the duration of the cycle. The specific activity of these wire segments can be used in conjunction with the presented cycle history information to determine the thermal- and fast-neutron fluence rates for each irradiation position during the cycle using the referenced standard test methods.
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Major accomplishments for the months of January, February and March 2020 in the fields of Fuels Development Qualification, High Temperature Materials Development, Graphite and GRC Modeling and Validation.
Report of activity for the Fiscal Year 2020 third quarter
Presented at Pittsburgh Coal Conference (virtual) September 8-11, 2020
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