Why does Uranium-238 absorb fast moving neutrons

  • Thread starter Thread starter Ed Aboud
  • Start date Start date
  • Tags Tags
    Neutrons
AI Thread Summary
Uranium-238 absorbs fast-moving neutrons effectively due to its nuclear properties, despite initial intuition suggesting otherwise. It captures both slow and fast neutrons, but the likelihood of fission in Uranium-235 is significantly higher with slow neutrons. This results in a more favorable ratio for fission in U-235 compared to neutron capture in U-238 when slow neutrons are involved. The discussion highlights the importance of the pairing force in understanding these interactions. Overall, the behavior of U-238 in neutron absorption is complex and influenced by various nuclear dynamics.
Ed Aboud
Messages
200
Reaction score
0
Hi all.
Why does Uranium-238 absorb fast moving neutrons very well but not slow moving ones?
My intuition tells me it should be the other way around but its not.
Thanks.
 
Engineering news on Phys.org
Hmm, I still don't really get it.
 
U-238 captures slow neutrons as well as fast, and even a bit more slow than fast. However, U-235 fissions MUCH MORE easily with slow neutrons than with fast, and hence the ratio for fission in U-235 over capture in U-238 is more favorable for slow neutrons than for fast.
 
Oh ok, I get it now. Thanks very much.
 
it has to with the pairing force, the answer vanesh gave you.
 

Thread 'MCNP6 BURN Card Issue: Zero Burnup for Material in Fuel Assembly'

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...
View full post »

Thread 'Beginner Seeking Help: MCNP6 Burnup Example (OECD-NEA Benchmark)'

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...
View full post »

Similar threads

What actually happens to the U-238 part of fuel in a nuclear reactor?
Replies
7
Views
3K
MCNP and simple nuclear physics
Replies
9
Views
3K
Fast reactors: How can they work without a moderator?
Replies
3
Views
3K
Simulation of Pu-238 Decay
Replies
2
Views
835
MCNP6: Energy and angular distribution of neutrons after moderation
Replies
6
Views
3K
Why Might Beryllium Be Considered for Use in Nuclear Reactor Fuel?
Replies
3
Views
2K
Which actinides are not fissile?
Replies
13
Views
9K
Doppler effect neutron absorption
Replies
1
Views
2K
FLUKA: Why Does Neutron Flux Increase When Passing Through a Moderator?
Replies
5
Views
3K
Neutron capture and fission reaction
Replies
15
Views
3K

Hot Threads

Recent Insights

Back
Top