Hi every body, i'm working and calculating the keff, and total

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SUMMARY

The discussion centers on calculating the effective multiplication factor (keff) and total fission produced during a critical accident in a cylindrical model using MCNP 5. Key calculations involve the variable multiply coefficient (keff) and the total fission, which can be determined through the integral of the fission cross-section (Ʃf) and neutron flux (∅(x)) over volume and time. The user seeks clarification on Ʃf, specifically for thermal neutrons impacting U-235, and the discussion provides resources for finding the necessary fission cross-section data.

PREREQUISITES
  • Understanding of neutron transport theory and criticality safety.
  • Familiarity with MCNP 5 for neutron transport simulations.
  • Knowledge of fission cross-sections and their significance in nuclear reactions.
  • Basic calculus for integrating functions over volume and time.
NEXT STEPS
  • Learn how to calculate fission cross-sections for U-235 using the formula Ʃf = NU235 * σf.
  • Research the integration of neutron flux over a volume to determine total fission produced.
  • Explore the MCNP 5 documentation for accessing cross-section libraries.
  • Investigate the effects of different material states (solid, ceramic, aqueous) on keff calculations.
USEFUL FOR

Nuclear engineers, researchers in criticality safety, and students studying neutron transport and fission processes will benefit from this discussion.

tiepngh
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hi every body, I'm working and calculating the keff, and total

hi every body, I'm working and calculating the keff, and total fission produced in critical accident in the cylindrical model. If critical mass is roached.
I calculated variable multiply coefficient (keff) by MCNP 5.
also total fission produced i don't know how to calculate this variable,
I really need your help.
thanks you very much for your help.
 
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tiepngh said:
hi every body, I'm working and calculating the keff, and total fission produced in critical accident in the cylindrical model. If critical mass is roached.
I calculated variable multiply coefficient (keff) by MCNP 5.
also total fission produced i don't know how to calculate this variable,
I really need your help.
thanks you very much for your help.

Well, you could just take the fuel depletion in atoms...
 


tiepngh said:
hi every body, I'm working and calculating the keff, and total fission produced in critical accident in the cylindrical model. If critical mass is roached.
I calculated variable multiply coefficient (keff) by MCNP 5.
also total fission produced i don't know how to calculate this variable,
I really need your help.
thanks you very much for your help.
Total fission would be the integral of Ʃf and flux ∅(x) over the volume and time. One could reasonably assume a Gaussian pulse and mostly a prompt neutron flux.
 


Astronuc said:
Total fission would be the integral of Ʃf and flux ∅(x) over the volume and time. One could reasonably assume a Gaussian pulse and mostly a prompt neutron flux.
thanks for your answer,
and could you show me about Ʃf clearly, eg i want to calculate with thermal neutron(E < 6.625e-7 MeV), and fission cause of U235. how is Ʃf?
Thanks you !
 


tiepngh said:
thanks for your answer,
and could you show me about Ʃf clearly, eg i want to calculate with thermal neutron(E < 6.625e-7 MeV), and fission cause of U235. how is Ʃf?
Thanks you !
What is the initial form of the material? Solid metal, ceramic, or a aqueous solution? One can estimate the reactivity from keff-1.

Ʃf = NU235 * σf

and N is the atomic density of the material.
 


Astronuc said:
What is the initial form of the material? Solid metal, ceramic, or a aqueous solution? One can estimate the reactivity from keff-1.

Ʃf = NU235 * σf

and N is the atomic density of the material.
Thanks you very much, i have calculate critical model with solution material, one things i don't know that σf, and could you show me how to find it?
Solution : UO2(NO3)2 . 6 H2O.
 


tiepngh said:
Thanks you very much, i have calculate critical model with solution material, one things i don't know that σf, and could you show me how to find it?
Solution : UO2(NO3)2 . 6 H2O.
I would expect it is in a library of cross-sections in MCNP 5.

Otherwise, http://www.nndc.bnl.gov/sigma/index.jsp?as=235&lib=endfb7.1&nsub=10 - and find (n,total fission).

There will be no delayed neutrons.
 

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