Nuclear data libraries in ACE format for MCNP5

AI Thread Summary
Continuous-energy neutron cross section libraries in ACE format for MCNP5, including endf66a, endf66b, endf66c, and misc5xs, are sought after, along with continuous-energy photon tables like mcplib and its variants. The OECD NEA Janis web portal is mentioned as a resource for evaluating nuclear data. Users are encouraged to share information on the availability of pre-processed ACE format libraries. The original poster successfully obtained all necessary ACE libraries from Oak Ridge RSICC. This highlights the importance of community support in accessing specialized nuclear data resources.
AlbeFerS
Messages
6
Reaction score
0
Hello everyone,

I need some continuous-energy neutron cross sections libraries in ACE format for MCNP5 (endf66a, endf66b, endf66c, misc5xs) as well as the continuous-energy photon tables (mcplib, mcplib02, mcplib03, mcplib04).

Here: http://www.oecd-nea.org/janisweb/search/endf , we have evaluated nuclear data but, does everyone know if there are libraries already processed in ACE format available for downloading?

Thank you in advance
 
Engineering news on Phys.org
Finally I got all the ACE libraries for MCNP5 from Oak Ridge RSICC.
 

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

Computation for nuclear reactor systems
Replies
1
Views
3K
Where can I find d-d reaction cross-section data for higher energies?
Replies
4
Views
3K
I need help please for Simulation of Xenon Oscillations in a Nuclear Reactor
Replies
4
Views
1K
Questions about Neutron scattering theory
Replies
4
Views
3K
Extracting thermal neutron cross-sections from ENDF libraries?
Replies
3
Views
4K
B Assistance finding the nuclear energy difference during nuclear fusion
Replies
5
Views
2K
Help Needed: Discrepancy in 2D Monte Carlo Neutron Transport Simulation
Replies
1
Views
96
I Problem calculating Li-7 neutron - induced nuclear reaction
Replies
10
Views
2K
I Status and Prospects of PADME Run 3
Replies
10
Views
3K
I Updates on the experiments X17
2
Replies
50
Views
9K

Hot Threads

Recent Insights

Back
Top