Are there any references/codes/methods which able to find such a relation/formula/theory between the neutron fluence/flux with burnup rate.
Thank you ~~
Maybe you can try OpenMC code developed by MIT, and the code is open source. However, the burnup calculation is difficult and the accuration can not be ensured sometimes.
Dear Almomani: as it is well known, neutron fluence is related to burn-up. After time integration of neutron flux, it is clear that neutron fluence depends also on the energy. This dependency with energy is a consequence of the geometry (or the system) under study. The fast neutron flux will depend on the type of nuclear reactor, then the neutron fluence. What you can do is, a simple estimation of fast neutron flux (at energy which you will decide, >0.1 MeV for example) and calculate this fluence as burn-up function. I also suggest you, check bibliografy about it at IAEA.
Concerning energy reactor the fast neutron fluence is increased with burnup because of neutron fluence is defined as the time integral of the neutron flux density.
If you have asked about fast neutron flux then it is increased also during ONE fuel cycle because of average energy of fission neutron by plutonium-239 is slighnly higher against uranium-235.
There is no law for fast neutron flux dependence with burnup during couple fuel cycles.
For the calculations one can use any Monte code with burnup.
This may also help a little
Article Spent fuel interrogation using delayed fast neutron spectrum...
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