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how to convert MNCP F4 tally to neutron flux (n/cm2s). and how to calculate average neutron produced per fission from given nuclear fuel mixture (eg. mixture of 80.8% enriched U-235 and U-238 as fuel material).
The conversion from MNCP F4 tally to neutron flux is a simple calculation using the following formula: Neutron Flux (n/cm2s) = MNCP F4 Tally / Detector Efficiency. The detector efficiency can be found in the detector's calibration certificate or can be determined through experimental measurements.
MNCP F4 tally is a type of neutron detector that uses Multi-Channel Analyzer (MCA) to record the energy spectrum of neutrons. It is commonly used in neutron flux measurements and neutron activation analysis.
Yes, there are other formulas and methods for converting MNCP F4 tally to neutron flux. However, the formula Neutron Flux (n/cm2s) = MNCP F4 Tally / Detector Efficiency is the most commonly used and recommended by experts.
MNCP F4 tally is typically measured in counts or counts per second. Neutron flux is measured in units of neutrons per square centimeter per second (n/cm2s).
Yes, converting MNCP F4 tally to neutron flux is necessary to accurately measure the neutron flux in a given area. The MNCP F4 tally alone does not provide a quantitative measurement of neutron flux, but rather a qualitative measure of neutron activity.