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

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The discussion focuses on calculating the effective multiplication factor (keff) and total fission produced in a critical accident scenario using a cylindrical model. The user has successfully calculated keff using MCNP 5 but seeks assistance with determining total fission produced. Key suggestions include integrating the fission cross-section (Ʃf) and neutron flux over volume and time, with specific emphasis on thermal neutrons and U235 fission. Additional guidance is provided on finding the fission cross-section (σf) from libraries or databases. The user expresses gratitude for the help and aims to resolve their calculations.
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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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