Gamma energies from fast neutron activation.

AI Thread Summary
Gamma energies from fast neutron activation are not well-documented, making it challenging for researchers to find reliable sources. The discussion highlights the need for characteristic gamma rays for MCNP projects, with users currently relying on self-generated data. Suggestions include utilizing the BNL-NNDC data, specifically the ADOPTED LEVELS, GAMMAS, or ENSDF files for specific isotopes. Participants are encouraged to seek insights from the MCNP team at LANL for additional resources. Accessing comprehensive libraries or databases is crucial for accurate gamma energy information in neutron activation studies.
theCandyman
Messages
397
Reaction score
2
I've looked around and not found any good sources on the gamma energies from fast neutron activation. I need to find the characteristic gamma rays for an MCNP project and currently, I'm relying on energies I get from running some simple cases I set up for the element of interest. Does anyone know of any source I can use?
 
Engineering news on Phys.org
Perhaps the MCNP folks at LANL have some ideas or know of a library.

One could try to use the BNL-NNDC data - perhaps ADOPTED LEVELS, GAMMAS or ENSDF files, e.g. http://www.nndc.bnl.gov/chart/getdataset.jsp?nucleus=65NI&unc=nds
 

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

How to use this MCNP physics model?
Replies
7
Views
1K
I want to see the peak for an explosive detector simulation in MCNP
Replies
1
Views
1K
How to Calculate the Source Activity from Dose Rate in MCBEND
Replies
0
Views
2K
MCNP6: Energy and angular distribution of neutrons after moderation
Replies
6
Views
3K
How can I model both photons and neutrons with MCNP?
Replies
7
Views
3K
Looking for information on gamma light sources
Replies
3
Views
2K
MCNP neutron energy deposition in water
Replies
1
Views
2K
Radiation effects/damage in structural alloys and ceramics
Replies
0
Views
1K
Atomic displacements, replacements, and radiation damage of structural materials
Replies
1
Views
643
FLUKA: Why Does Neutron Flux Increase When Passing Through a Moderator?
Replies
5
Views
3K

Hot Threads

Recent Insights

Back
Top