# How to convert MNCP F4 tally to neutron flux (n/cm2s).

• ISAAC BAIDOO
In summary, to convert MNCP F4 tally to neutron flux (n/cm2s), you need to use a conversion factor which is the average number of neutrons produced per fission divided by the total number of fission events. The average number of neutrons produced per fission can be calculated from the given nuclear fuel mixture using the formula provided. Once you have the total number of fission events and the average number of neutrons produced per fission, you can calculate the neutron flux.
ISAAC BAIDOO
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).

To convert MNCP F4 tally to neutron flux (n/cm2s), you will need to use a conversion factor. This factor is the average number of neutrons produced per fission divided by the total number of fission events. The total number of fission events can be found from the MNCP F4 tally. The average number of neutrons produced per fission can be calculated from the given nuclear fuel mixture by using the following equation:Average neutrons per fission = (0.808 x 2.43) + (0.192 x 2.88)Where 0.808 and 0.192 represent the percentage of U-235 and U-238 in the fuel mixture, respectively, and 2.43 and 2.88 represent the average number of neutrons produced per fission for U-235 and U-238, respectively. Therefore, the average number of neutrons produced per fission in the given nuclear fuel mixture is 2.67 neutrons per fission. Once you have the total number of fission events from the MNCP F4 tally and the average number of neutrons produced per fission, you can then calculate the neutron flux (n/cm2s) by dividing the total number of fission events by the average number of neutrons produced per fission.

Hi there,

Converting MNCP F4 tally to neutron flux can be done by using the following formula: flux = tally / area / time. The MNCP F4 tally is a measure of the number of neutrons produced per unit area and time, so by dividing it by the area and time, you can get the neutron flux.

As for calculating the average neutron produced per fission, you can use the following formula: average neutron per fission = (number of neutrons produced by U-235 * enrichment percentage) + (number of neutrons produced by U-238 * (1-enrichment percentage)). In this case, the number of neutrons produced by U-235 is 2.43 and the number of neutrons produced by U-238 is 2.07. So for a fuel mixture of 80.8% enriched U-235 and 19.2% U-238, the average neutron produced per fission would be (2.43 * 0.808) + (2.07 * 0.192) = 2.28.

Hope this helps!

## 1. How do I convert MNCP F4 tally to neutron flux?

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.

## 2. What is MNCP F4 tally?

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.

## 3. Can I convert MNCP F4 tally to neutron flux using a different formula?

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.

## 4. What unit is used for MNCP F4 tally and neutron flux?

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).

## 5. Is it necessary to convert MNCP F4 tally to neutron flux?

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.

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