Resonance Self Shielding Correction Definition

AI Thread Summary
Resonance self-shielding correction refers to the phenomenon in heterogeneous reactors where fuel particles reduce their exposure to neutron flux due to their own resonance absorption cross-section. In such reactors, neutrons can slow down in the moderator without being absorbed by fuel atoms, leading to a lower neutron flux in the fuel pins. The correction factor quantifies the extent of this self-shielding effect, which is influenced by the reactor's geometry and composition. There is some confusion in the community regarding the definitions of self-shielding and resonance self-shielding, as they are closely related concepts. Understanding this correction is crucial for accurate reactor modeling and neutron flux calculations.
terryphi
Messages
57
Reaction score
0
Hello,

I was hoping that someone might be able to provide a definition for the "resonance self shielding correction".

-TP
 
Engineering news on Phys.org
In a homogenous reactor (e.g. liquid fuel mixed with moderator), the neutron flux is relatively constant because there is a equal distribution of fuel and moderator particles. When you have a heterogeneous reactor (e.g. solid fuel pins surrounded by water), neutrons have a higher chance of slowing down without being resonance absorbed because they can bounce around in the moderator without worrying about hitting a fuel atom while in the resonance range. The fuel shields itself from the neutron flux due to its resonance absorption cross section and therefore there is a relatively lower neutron flux in the fuel pins compared to the moderator.

The self-shielding correction factor is the degree of this effect, which mainly depends on the geometry and composition of the reactor.
 
I believe it refers to the high cross-section resonances shielding themselves and other resonances. The self-shielding is the same concept more or less.
 
I've actually heard both definitions, so I'm confused.
 
Hello everyone, I am currently working on a burnup calculation for a fuel assembly with repeated geometric structures using MCNP6. I have defined two materials (Material 1 and Material 2) which are actually the same material but located in different positions. However, after running the calculation with the BURN card, I am encountering an issue where all burnup information(power fraction(Initial input is 1,but output file is 0), burnup, mass, etc.) for Material 2 is zero, while Material 1...
Hi everyone, I'm a complete beginner with MCNP and trying to learn how to perform burnup calculations. Right now, I'm feeling a bit lost and not sure where to start. I found the OECD-NEA Burnup Credit Calculational Criticality Benchmark (Phase I-B) and was wondering if anyone has worked through this specific benchmark using MCNP6? If so, would you be willing to share your MCNP input file for it? Seeing an actual working example would be incredibly helpful for my learning. I'd be really...
Back
Top